Raising the JET maximum toroidal field from 3.45 to 4.0 T results in increased forces and stresses in the key machine components during normal operations and disruptions. These forces and stresses have been predicted for 4.0 T operation and compared with allowable values derived from tests on used and spare JET coils. In this assessment significant lessons related to the design, manufacture and operation of key Tokamak components have also been learnt.

The JET Toroidal Field (TF) coils were originally designed for operation at 3.4 tesla. In order to upgrade the field to 4 tesla and thus improve the performance of the JET machine, new mechanical tests and analysis were carried out on the insulation of TF coil samples. They are aimed at investigating the mechanical properties and the status of the insulation in order to set allowable stresses and force limits. In particular since the shear stress in the insulation is strongly affected by the shear modulus of elasticity G, it is important to measure this parameter. A method for the measurement of G in glass-resin fibres, the V-notched beam method (Iosipescu method), was applied. The particular shape of the rectangular Iosipescu V-notched sample and the particular modality of force application produce pure shear stress for a reliable measurement of the G value and of the shear strength of the insulation. The effect of temperature on these mechanical properties was also investigated. Results show higher average shear strength with lower scatter compared with previous tests on conventional rectangular samples, thus confirming the reliability of the method. Micrographic analysis of the insulation and comparison between the straight and curved regions of the magnet, where the highest stress occurs, confirm the good quality of the impregnation of the coil. Glass-resin content, void content, micros and TG measurements have been performed on different samples and correlation between the different properties of the insulation investigated. Moreover fatigue tests at different temperatures were performed and data analyzed with the cumulative damage technique, which allows for an extrapolation of the fatigue curve with less samples than the standard method.

The toroidal magnetic field coils on JET were designed to operate at 3.4 T During the JET tritium campaign (DTE1, 1997) similar to 150 pulses were run at 3.8 T. Recently operation at 4.0 T has been commissioned and used briefly in plasma operation. An extensive analytical and experimental study over the last four years has concluded that limited operation at 4.0 Tin a wide range of scenarios is acceptable.

Recent experiments in D-T plasmas on the JET and TFTR tokamaks have evaluated a wide range of ITER relevant ion cyclotron heating scenarios. Absorption of fast waves at the second-harmonic tritium resonance has provided bulk ion heating in TFTR supershots and electron heating in JET H-mode discharges. In JET, deuterium minority heating has generated 1.7 MW of fusion power with 6 MW of radio frequency power giving a record steady-state Q-value of 0.22. Strong bulk ion heating has been achieved with He-3 minority heating with central ion temperatures up to 13 keV being produced in H-modes with a density of 3.6 x 10(19) m(-3) Hydrogen, deuterium and He-3 minority heating methods have produced plasmas with normalized confinement times greater than or equal to that required by ITER for ignition. These H-modes are characterized by small-amplitude, high-frequency ELMs, each of which transports less than 1.5% of the plasma energy content to the limiters. The heavy minority scheme of tritium in a deuterium plasma has been demonstrated both as a heating scheme and a generator of suprathermal neutrons. On TFTR mode conversion to an ion Bernstein wave has achieved central bulk ion heating in supershots with target ion temperatures greater than 20 keV.

The design of the Joint European Torus was conceived with inherent flexibility to accommodate modifications and upgradings to match the evolving requirements of the physics programme, while preserving basic machine structure. The first major upgrading was to increase the plasma current capability from 4.8 to 7 MA in limiter configuration and from 3.0 to 5 MA in X-point configuration. The second change was the progressive covering of the vessel walls with low-Z materials such as graphite and beryllium. The most recent major modification was to make JET into a pumped divertor machine. Three diverters are being tested in sequence (Mark I, II, IIGB), in support of the ITER design. JET is operating at present with Mark II both in D-D and in D-T Thus, the installation of Mark IIGB will be performed using only remote handling techniques. Divertor plasmas are more vertically unstable, and so a new plasma control system had to be designed and implemented. The engineering instrumentation of the machine has been upgraded, for machine protection and to monitor and study new phenomena such as sideways vessel displacements, caused by plasma disruptions. An in depth reassessment of the toroidal coils, of the mechanical structure and of the vessel is in progress. This includes finite element calculations and mechanical tests on samples and on two toroidal field whole coils, to evaluate the machine capability to operate at higher toroidal field (from 3.4 T to 4.0 T) and operation at 3.8 T has been undertaken already. In the early phase of the 1997 D-T campaign /spl sim/13 MW of fusion power have been produced.

Two pumped diverters have been installed and tested in JET under ITER relevant conditions. A closed divertor is found to increase the particle and impurity exhaust rate in agreement with code modelling. Excellent power handling is demonstrated, allowing high current discharges with record stored energy (15 MJ) and quasi-steady-state discharges with high fusion triple product (4 x 10(20) m(-3) s keV). The ITERH93-P confinement scaling law is confirmed over abroad range and a more favourable beta scaling is found. No hysteresis is found in the II-mode power threshold. A data base of highly radiating discharges including impurity seeding describes the relationship between radiated power, impurity concentration and density. First results of high performance with optimised magnetic shear are presented. Strong internal confinement barriers develop despite the relatively small input of toroidal momentum and particle fuelling from ICRH and NBI.

The inherent flexibility of JET's original concept has permitted several engineering upgradings and modifications, to address a large variety of plasma and fusion physics issues. The most recent major modification has been the installation of an axisymmetric single-null pumped divertor (Mark I), successfully operated in the experimental period 1994–1995. Following the divertor optimization programme a new, more closed, divertor configuration has now been installed (Mark II), which has shown a better power handling capability and substantially improved neutral particle retention. A key feature of the new design is the possibility to replace the divertor target plate structure using full remote handling techniques following extended D-T operations. Toroidal asymmetries of vessel forces due to Vertical Displacement Events (VDE) and halo currents were experienced since 1994, leading in some cases to sideways movements of the vessel of 7 mm. This has required modification and upgrading of the vacuum vessel support system. Gap control of plasma position and shape and machine protection systems have been developed further, leading to increased experimental availability. Future development foresees the installation of a Mark II Gas Box divertor structure, while studies are underway to increase the toroidal field capability from 3.45 to 4 T and the additional heating power by increasing the NB injector output from 80 kV, 60 A to 120 kV, 60 A and by using wide band matching for ICRF.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

This paper presents an overview of results of the 1994/95 experimental campaign on JET with the new pumped divertor and draws implications for ITER in the areas of detached and radiative divertor plasmas, the use of beryllium as a divertor target the material, the confinement properties of discharges with the same dimensionless parameters (except for the dimensionless Larmor radius) as ITER and the effect of varying the toroidal magnetic field ripple in the ITER relevant range. Discharges with high fusion performance at high current, in steady-state with ELMs and in the ELM-free hot-ion H-mode, are also reported. Limits to operations are discussed and projections to D-T performance are made.

Since 1994 the JET experiment has been operated with a divertor, with currents up to 6MA. Disruptions are generally accompanied by vertical plasma displacements giving rise to vertical forces at the torus. Vertical force swings up to 5MN were recorded at vessel supports. The forces are toroidally non-uniform, with peaking factors up to 1.8. Global sideways displacements of the torus, up to about 5 mm, were also recorded in a number of disruptions. They are interpreted as consequence of a large amplitude m=1, n=1 kink mode. Disruptions led to damage of some components inside and also outside the vessel, such as internal saddle coils, and beryllium evaporator heads.

An overview of the present design of the ITER coil power supply system is presented, along with brief descriptions of the main components including the AC/DC converters, switching networks, and discharge circuits.

The most recent modification to JET has been the installation of a single-null pumped divertor, for active control of plasma impurities. This is to address central physics issues relevant to the design of a `next step' tokamak. Experiments conducted during the 1994-95 campaign, with plasma currents up to 6MA, have shown that the Mark I divertor, which makes use of strike point sweeping across the target plates, is a suitable tool to control the influx of impurities in the plasma core. The operation of a tokamak with a pumped divertor has been characterised in detail. However the divertor configuration must be optimised to better meet ITER requirements. Therefore an improved (more closed) divertor structure, which may not require sweeping, is under assembly at present (Mark II). It is designed, in addition, to allow divertor tile structures to be fully replaceable by remote handling techniques, following D-T fusion experiments. New types of events involving electromechanical interactions of plasma with the vessel and in-vessel structural components have been encountered, due to plasma vertical instabilities and disruptions (such as toroidal asymmetries of vacuum vessel forces and side-ways vessel displacements). The physics and engineering experimental work performed in JET is primarily dedicated to the finalisation of the ITER design

The JET neutral beam injection system has proved to be both effective and reliable as a plasma heating device. The ion heating and plasma fuelling characteristics of the system are ideally suited to the production of high fusion performance plasmas while the flexibility in the choice of beam species (H, D, T, 3He or 4He) and the ability to inject into almost any JET plasma configuration allows a wide variety of related physics experiments to be carried out. The capability to inject (for the first time) tritium beams was essential to the successful execution of the first tritium experiments in which 1.7 MW of power from DT fusion reactions was generated.

Eleven years of JET operations have brought studies of tokamak plasmas up to reactor-relevant performance. This has resolved some key issues for the design of a ''next step'' machine. However it has been clearly shown that active control of the influx of impurities is essential for a long-burn fusion reactor. Therefore, a single-null pumped divertor has been installed inside the JET vacuum Vessel to study impurity control, and it is now operational. It consists of four poloidal coils (manufactured inside the vessel), carbon-fibre composite (CFC) target plates and a toroidal cryo-pump. The new magnetic configuration produces plasma shapes which required a complete redesign of the first wall, including r.f. antennae and limiters. New power supplies were introduced for the divertor coils and for plasma fast vertical position control. Additional cooling systems, a new digital plasma control, a comprehensive machine protection system and new diagnostics to measure plasma parameters in the divertor region were also installed. Well-controlled 4-MA discharges have already been achieved.

The European Union ITER Home Team analysed the possibility of feeding ITER directly from the 400 kV grid. The study was based on the assumption that ITER will be located on the sites where the 400 kV grid feeds fusion experiments in Europe at present, and was performed in cooperation with the Electric Companies concerned. This study shows, as a first approximation, the ability of the European Grid to supply the ITER load, provided that measures such as increasing the power reserve and providing reactive power compensation are taken as the site location demands

The concept and the key features of the Joint European Torus (JET), flagship of the integrated European fusion research programme, departed considerably from those of other large tokamaks under design in the early 1970s. D-shape toroidal coils and vacuum vessel and large volume high current plasma were unique and controversial features of JET. Moreover, since the early phase of the JET design, due consideration was given to D-T operations, including remote handling capability and D-T compatible peripheral systems. JET experimental results have confirmed the validity of these design choices. In turn, these choices had an impact on the development of other tokamaks and of ‘next step’ design proposals, such as NET (Next European Torus) and later ITER (International Tokamak Experimental Reactor). Experimental evidence from JET and other tokamaks has clearly shown that the control of impurities is a key issue for finalizing ITER design. Therefore, the present JET programme is focussed on a divertor programme with thermonuclear grade plasmas tailored to ITER needs and to the exploration of advanced tokamak concepts for further enhanced global performance.

Four coils have been built and installed in the JET vacuum vessel to produce divertor plasmas. The coils are copper with glass epoxy insulation and are enclosed in vacuum tight Inconel cases. At the coil contractor's factory, the coil parts were manufactured and process techniques qualified. In the JET vacuum vessel the conductors bars were brazed to form the coils, which were inserted in the casings and impregnated and cured with epoxy resin

Three of the toroidal field (TF) coils of the JET tokamak have developed interturn faults. These faults have not been catastrophic and it has been possible to continue operation and replace the faulty coils with spares at the next pre-planned shutdown. The faults were found to be due to water leaks. The coil coolant was changed from water to an insulating fluid. All known faulty coils have been changed and latest measurements do not detect any faults on the installed coils