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The experimental programme of Tore Supra, the largest superconducting tokamak in the world (a = 0.72 m, R = 2.4 m, Ip < 1.7 MA, BT < 4.5 T) was devoted in 2003 to studying the heat removal capability and particle exhaust in steady-state fully non-inductive current drive discharges simultaneously. This required both advanced technology integration and steady-state real-time plasma control. In particular, an improvement of the plasma position to within a range of few millimetres, and new real-time controls of radio frequency power and various actuators built around a shared memory network, have allowed Tore Supra to access a powerful steady-state regime with an improved safety level for the actively cooled plasma facing components. Feedback controlled fully non-inductive plasma discharges have been sustained in a steady-state regime for up to 6 min with a new world record for injected–extracted energy exceeding 1 GJ. Experimental results and an analysis of the physics involved in these discharges are reported and discussed.

The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, viz. ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is ∼ 10-20 %. This is larger than the ~3-4% deduced from post mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes were developed.

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In the perspective of operating tungsten monoblocks in WEST, the ongoing major upgrade of the Tore Supra tokamak, a dedicated modelling effort has been carried out to simulate the interaction between the edge plasma and the tungsten wall. A new transport code, SolEdge2D-EIRENE, has been developed with the ability to simulate the plasma up to the first wall. This is especially important for steady state operation, where thermal loads on all the plasma facing components, even remote from the plasma, are of interest. Moreover, main chamber tungsten sources are thought to dominate the contamination of the plasma core. We present here in particular new developments aimed at improving the description of the interface between the plasma and the wall, namely a way to treat sheath physics in a more faithful way using the output of 1D particle in cell simulations. Moreover, different models for prompt redeposition have been implemented and are compared. The latter is shown to play an important role in the balance between divertor and main chamber sources.

Disruptions are a major threat for future tokamaks, including ITER. Disruption-generated heat loads, electromagnetic forces and runaway electrons will not be tolerable for next-generation devices. Massive noble gas injection is foreseen as a standard mitigation system for these tokamaks. Disruption mitigation experiments have been carried out on Tore Supra to study various injection scenarios and to investigate gas jet penetration and mixing. Comparisons of different gases (He, Ne, Ar, He/Ar mixture) and amounts (from 5 to 500 Pa m3) were made, showing that light gases are more efficient regarding runaway electron suppression than heavier gases. Eddy currents in the limiter are moderately reduced by all the gases, and may be more dependent on the time constants of the structures than on the gas species. The density rise induced by the massive injection before the thermal quench is higher and faster with light gases. Gas jet penetration in the cooling phase is observed to be shallow and independent of the gas nature and amount. The gas cold front is stopped along the q = 2 surface where it triggers MHD instabilities, expelling thermal energy from the plasma core.

Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side of the 3 He ion cyclotron resonance layer in D and 4 He plasmas and subsequently damped on the bulk electrons. The resulting electron power deposition, measured using ICRF power modulation, is narrow with a typical full-width at half-maximum of ≈30 cm (i.e. about 30% of the minor radius) and the total deposited power to electrons comprises at least up to 80% of the applied ICRF power. The ICRF mode conversion power deposition has been kept constant using 3 He bleed throughout the ICRF phase with a typical duration of 4-6 s, i.e. 15-40 energy confinement times. Using waves propagating in the counter-current direction minimizes competing ion damping in the presence of co-injected deuterium beam ions.

In the recent JET experimental campaigns with the new ITER-like wall (JET-ILW), major progress has been achieved in the characterization and operation of the H-mode regime in metallic environments: (i) plasma breakdown has been achieved at the first attempt and X-point L-mode operation recovered in a few days of operation; (ii) stationary and stable type-I ELMy H-modes with βN ∼ 1.4 have been achieved in low and high triangularity ITER-like shape plasmas and are showing that their operational domain at H = 1 is significantly reduced with the JET-ILW mainly because of the need to inject a large amount of gas (above 1022 D s−1) to control core radiation; (iii) in contrast, the hybrid H-mode scenario has reached an H factor of 1.2–1.3 at βN of 3 for 2–3 s; and, (iv) in comparison to carbon equivalent discharges, total radiation is similar but the edge radiation is lower and Zeff of the order of 1.3–1.4. Strong core radiation peaking is observed in H-mode discharges at a low gas fuelling rate (i.e. below 0.5 × 1022 D s−1) and low ELM frequency (typically less than 10 Hz), even when the tungsten influx from the diverter is constant. High-Z impurity transport from the plasma edge to the core appears to be the dominant factor to explain these observations. This paper reviews the major physics and operational achievements and challenges that an ITER-like wall configuration has to face to produce stable plasma scenarios with maximized performance.