[en] Full text: The scientific capabilities of the JET tokamak have recently been enhanced with the successful installation of a modified divertor allowing high triangularity ITER relevant scenarios. In addition, up to 17 new or upgraded diagnostics are foreseen to be operational for the Experimental Campaigns in 2005/2006. Moreover, a new ITER-like Ion Cyclotron Resonant Heating (ICRH) antenna will be installed in 2006. Besides bringing the new systems to full performance, the main objectives of the Experimental Campaigns 2005/2006 will be: to prepare ITER operating scenarios, to address specific physics issues of direct relevance to ITER (e.g. transport physics, burning plasma physics) as well as to study critical issues potentially impacting the detailed design of ITER components (e.g. first wall, heating and current drive systems, diagnostic). For the longer term, work on a 'JET programme in support of ITER' has begun, which aims at making optimal use of JET's unique features: physical size, and the capability to handle beryllium (Be) and tritium (T), following a further set of major enhancements, foreseen to be implemented in 2008. ITER is currently designed to have a beryllium-clad first wall, tungsten (W) brushes at the divertor entrance and carbon fibre reinforced carbon (CFC) tiles at the divertor strike points. This materials combination has never been tested in a tokamak. A modification of the JET wall materials is planned to bring operational experience in steady and transient conditions with ITER first wall and divertor materials, relevant geometry and relevant plasma parameters. Such a programme could provide input to optimise the choice of first wall materials on ITER, to balance tritium retention and configuration flexibility. Furthermore, the compatibility of operating scenarios with the first wall would be tested at ITER-relevant parameters, especially when combined with the foreseen power upgrade. As the reference option in the frame of the ITER-like first wall project on JET, a combination of Be on the first wall and W-coated CFC tiles in the divertor is planned. This would provide experimental information on a C-free tokamak operation. An alternative design contains a Be first wall, t the divertor strike points and W on divertor baffles and dome, i.e. the configuration foreseen for the start of ITER. The decision on the pre erred installation scheme will be made in 2006, depending on ITER needs and the outcome of ongoing researches and developments on W-coating and bulk W technology. The installation is planned to be done during a year-long shutdown in 2008. This will be followed by a scientific programme in which critical issues for ITER will be addressed, including the control of T-retention, material erosion and migration, mixed materials effects, melt layer behaviour and impurity control, and operational scenario compatibility with a Be/W material mix. Additional neutral beam heating power, up to 36 MW for 20 s (compared to 25 MW for 10 s presently) will be provided by upgrading an existing beam box and power supplies for operation at higher current. Phasing two beam boxes will make it possible to deliver 20 MW of neutral beam power for up to 40 s, for full exploitation of the pulse length capability of the JET machine. This will be key to progress, in particular, hybrid and advanced scenarios, which require full or partial current profile control. Another enhancement will consist of the installation of a pellet injector with a high repetition rate (up to 50 Hz), for ELM pacemaking as demonstrated on ASDEX Upgrade, and to provide a reliable high fuelling rate capability. In order to exploit the JET capability to provide ITER relevant plasma shapes at high plasma current, enhancements of plasma con rol are also foreseen. Firstly, upgrades of the radial field amplifier for vertical stabilisation, the vertical stabilisation controller and vertical speed diagnostic will (at least) double the survivable ELM size, and thereby permit the study of Type I ELM H modes at lower collisionality, and stronger shaping. Secondly, an increase of the number of power supplies for the external poloidal field coils will exploit the full capability of these coils resulting in improved boundary control for both single-null and near-double-null configurations, and a higher maximum plasma current for highly shaped configurations. Furthermore, diagnostics required to support further developments of ITER plasma scenarios, as well as ITER diagnostics that need to be tested on JET, will be implemented by upgrading existing systems. A study to install a set of ergodisation coils on JET for ELM control is also underway, in light of successful tests of a similar system on DIII-D. Along with these major projects several refurbishments are being prepared to ensure reliable operation of the JET machine in the years ahead