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At the 21st IAEA Fusion Energy Conference, 68 and 67 papers were presented in the categories of ITER activities and fusion technology, respectively. ITER performance prediction, results of technology R&D and the construction preparation provide good confidence in ITER realization. The superconducting tokamak EAST achieved the first plasma just before the conference. The construction of other new experimental machines has also shown steady progress. Future reactor studies stress the importance of down sizing and a steady-state approach. Reactor technology in the field of blanket including the ITER TBM programme and materials for the demonstration power plant showed sound progress in both R&D and design activities.

This paper summarizes the results presented at the 21st IAEA Fusion Energy Conference in the areas of innovative confinement concepts, magnetic confinement experiments and performance.

An overview is given of recent experimental results in the areas of divertor and exhaust, stability and waves and fast particles presented at the 2006 IAEA Fusion Energy Conference. Important new findings are presented from fusion devices worldwide, with a strong focus towards the needs of the major next step fusion experiment ITER.

Summary of the papers presented at the 21st IAEA Fusion Energy Conference (Chengdu, October 2006) and devoted to magnetic confinement theory and modelling.

Significant progress has been made on the DIII-D tokamak in the capability to control key plasma features and using such control to expand the operational limits of stationary and steady-state tokamak operation. Recent experiments have demonstrated the capability to suppress several key plasma instabilities of concern for ITER, including edge localized modes, neoclassical tearing modes and resistive wall modes. In addition, the ability to regulate the rotation and current density profiles through feedback control has been demonstrated. The use of these control techniques has allowed an expansion of the envelope of viable, stationary tokamak operation, highlighted by the demonstration of sustained (∼2 s) operation of β_N ≃ 4 (50% above the no-wall stability limit) as well as fully noninductive operation with β ≃ 3.5%. This development is supported by a vigorous basic fusion science programme, which has provided new insights into turbulence dynamics over a large range in spatial scales, new measurements of the structure of fast-ion instabilities and their effect on the fast-ion population and important information on the transport of carbon and associated tritium co-deposition on plasma facing surfaces.

Recent JT-60U experimental results towards the development of a steady-state advanced tokamak scenario are presented. The JT-60U experiments within the past two years emphasize the extension of the operation regimes and the development of active control for high bootstrap current fraction plasmas with a strong linkage between pressure and current profiles. Reduction of a toroidal magnetic field ripple by installing ferritic steel tiles decreases the fast ion loss and consequently enables one to access the new regimes due to an increase in net heating power. The decrease in the fast ion loss also contributes to access to the new regimes through confinement improvement attributed to the reduction of counter toroidal rotation. High β_N (∼4.2) exceeding the no-wall ideal limit is achieved at l_i = 0.8–1 in the high β_p ELMy H-mode plasmas with a weak positive shear at the large volume configuration close to the wall. The large volume configuration had so far suffered from the large toroidal field ripple. The value of β_N reaches the ideal wall limit, where RWM is suppressed by the plasma rotation. Small critical rotation velocity about 0.3% of Alfvén velocity is found for suppressing RWM. High β_N · HH_98(y,2) of 2.2 with β_N ∼ 2.3 and HH_98(y,2) ∼ 1 is sustained for 23.1 s, significantly longer than the current diffusion time (∼12τ_R) in the high β_p ELMy H-mode plasmas. This duration is extended to a time scale for enhancement of recycling owing to a decrease in the wall-pumping rate. Successful recycling and density control is demonstrated for ∼30 s in the enhanced recycling regime even with outgas from the wall by maximizing the divertor pumping rate in the high-density ELMy H-mode plasmas. With small effects of the toroidal field ripple reduction, the controllability is investigated in the reversed shear (RS) plasmas with a high bootstrap current fraction (f_BS) of ∼0.7–1. It is difficult to sustain these plasmas using a feedback control scheme for global parameter such as stored energy and the duration is determined by the ideal limit or slow ITB degradation in the linkage of current and pressure profiles. Real time q_min control is demonstrated with MSE diagnostic and LHCD at f_BS = 0.46 with the strong pressure and current profile linkage. Physics studies also progress towards burning plasma control. Semi-global change of the ITB structure is observed in both high β_p mode and RS plasmas. Complete NTM suppression is demonstrated with misalignment of ECCD position less than about half of the island width. Fast ELM propagation to the main chamber wall is observed, and significant hydrogen retention in the main chamber wall is found. In the plasma-shadowed area underneath the divertor region, the averaged carbon deposition rate is significantly low, however the (H + D)/C retention rate is high (∼0.8).

Tore Supra routinely addresses the physics and technology of very long-duration plasma discharges, thus bringing key information on critical issues of long pulse/steady-state operation of ITER. During the last two years, its scientific programme has entered a new phase towards higher power long pulse. Considerable progress has been accomplished towards reliable and safe operation at multi-MW power level in an actively cooled plasma facing components (PFCs) environment. Discharges lasting several resistive times, with combined lower hybrid current drive and ion cyclotron resonant heating at high power level, in plasmas close to the Greenwald limit and T_i close to T_e have been obtained. Improved confinement has been maintained together with q(0) ∼ 1.5 for durations up to 20 s. An advanced scheme for the physics and operational integration of plasma scenarios has been developed, and discharges at 7 MW total injected power level have been controlled in an MHD stable region for 60 s. The unique capability of the reflectometry system of Tore Supra has allowed detailed studies of particle/energy transport, and a weak dependence of confinement with β is corroborated by fluctuation measurements. A new mechanism is proposed to explain the D retention observed in long pulses. Analysis of large parallel flows in the scrape off layer (SOL) lends credence to the hypothesis that the SOL is fed by long range transport events that are created near the outboard midplane. Finally, new technological developments for in situ follow-up of PFCs ageing are reported.

Essential progress has been made in understanding the links between energy and particle transport in T-10 plasma with electron cyclotron heating (ECH) and turbulence behaviour. The interplay between modification of energy and particle confinement with plasma density increase and turbulence properties related to ion temperature gradient and trapped electron modes has been studied. Degradation of the energy confinement with an increase in ECH power coincides with the growth of the turbulence level in the low magnetic field side, while it has been found to be almost unchanged in the high magnetic field side. Operating mode with improved confinement has been obtained by repetitive deuterium pellet injection. Fast transport and MHD events triggered by pellet injection have been observed. The formation of electron internal transport during the current ramp-up phase with ECH is under investigation. Electron cyclotron current drive localized near q = 1 surface has been found to be effective for sawtooth control, and optimization of the profile of the driven current has been made.

Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower hybrid current drive (LHCD) experiments (P_LH < 900 kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at I_p ∼ 1.0 MA with good efficiency, I_drive = 0.4 P_LH/n_eoR (MA, MW, 10²⁰ m⁻³,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (B_th/B ∝ R^0.68±0.19) is inferred from its measured variation with B_T that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfvén cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as , and show a significant dependence on X-point topology. Fast camera images of intermittent turbulent structures at the plasma edge show they travel coherently through the SOL with a broad radial velocity distribution having a peak at about 1% of the ion sound speed, in qualitative agreement with theoretical models. Fast D_α diagnostics during gas puff imaging show a complex behaviour of discrete ELMs, starting with an n ≈ 10 precursor oscillation followed by a rapid primary ejection as the pedestal crashes and then multiple, slower secondary ejections.

Steady internal transport barriers (ITBs) are obtained in FTU at ITER-relevant magnetic field and density (n_e0 ≥ 1.3 × 10²⁰ m⁻³) in almost full non-inductive discharges, sustained by lower hybrid (LH) and electron cyclotron (EC) RF waves sources. Similarly to ITER, only electrons are directly heated which in turn heat ions via collisions and no momentum is injected. Collisions do not affect the mechanisms of turbulence suppression and energy transport. At the highest densities the ion thermal conductivity remains ≤ the ohmic level, while the energy confinement time exceeds the ITER 97-L scaling by about 1.6 times. The ITB radius can be varied in the range 0.2 ≤ r/a ≤ 0.65 modifying the radial profile of the LH driven current, acting mainly on the safety factor q. A liquid lithium limiter (LLL) of innovative design, composed of a mesh of porous capillaries, has been tested successfully for the first time on a medium size tokamak. The LLL surface showed no damage up to the maximum thermal load of 5 MW m⁻². With LLL cleaner plasmas are obtained and the particle recycling strongly drops; new interesting regimes of particle transport arise at high density, with highly peaked profiles. Significant progress in disruption mitigation by means of EC power has shown that they can be avoided when absorption occurs directly on the MHD islands driving the disruption. Feedback control/suppression of MHD tearing modes (TM, m = 2) with EC waves has been achieved relying on a real-time detection of the TM and of its radial location. Testing the collective Thomson scattering in ITER-relevant configuration has stressed that avoiding backscattered radiation to the source is very crucial. The theory of the evolution of fishbone-like instabilities driven by LH generated supra-thermal electrons in FTU is outlined, and its relation to the trapped α particles dynamics is stressed.

During the last two years on ASDEX Upgrade the physics base for ITER operation was significantly extended in both the foreseen standard H-mode scenario as well as the stationary improved H-mode scenario with enhanced performance. For anomalous transport (energy, particles, toroidal momentum) a multi-faceted picture of mode dominance in different plasma parameter regimes of ITG, TEM and ETG turbulence is evolving based on detailed measurements including fluctuation reflectometry and gyrokinetic calculations. Similar structures were found for natural and pellet induced, mitigated type I-ELMs. They develop to outward drifting helical filaments which are seen as footprints on plasma phasing components. New insights were gained on the interaction of energetic particles (driven by NBI and ICRH) with large scale instabilities (TAEs, NTMs, ELMs) based on new diagnostic and theoretical tools. The unexpected broadening of NBI driven currents beyond a certain heating power can be explained by a fast particle diffusion driven by small-scale turbulence. The active control of MHD instabilities (sawteeth, NTMs) concentrates on ECCD as proposed for ITER. NTMs were completely stabilized with very localized deposition of dc ECCD in improved H-modes, while for the deposition widths larger than the marginal island size as in ITER modulated injection phased with the island O-point was demonstrated to be advantageous. The tungsten coverage of the vessel interior was further extended up to 85%, where the highest erosion occurs at the LFS poloidal limiters, where it is dominated by ICRF accelerated impurities and fast particles from NBI. The W concentration can be kept at an acceptable level by ELM pace-making (using pellets) and by tailored central heating. The stationary improved H-mode extends ITER operation beyond its standard H-mode performance for confinement and beta with up to 0.4 (ITER reference value 0.2). Based on ASDEX Upgrade kinetic profiles it promises either a fusion performance up to ignition at full current or long pulses of up to 1 h at reduced current (hybrid scenario). This operation mode extends well into the ITER dimensionless parameter range except for ρ*. Besides the development of discharge scenarios, the influence of the edge pedestal on confinement and of the flat central q-profile on MHD and confinement was documented.

Recent experiment results from the HL-2A tokamak are presented in this paper. Supersonic molecular beam injection (SMBI) with liquid nitrogen temperature propellant is used. Low temperature SMBI can form hydrogen clusters that penetrate into the plasma more deeply and efficiently. Particle diffusion coefficient and convection velocity (D = 0.5–1.5 m² s⁻¹ and V_conv < 40 m s⁻¹, respectively) are obtained at the plasma periphery using modulated SMBI. Multi-probe measurements reveal the m = 0–1, n = 0 symmetries of directly measured low frequency (7–9 kHz) electric potential and field are simultaneously observed for the first time. Impurity transport is determined with the laser blow-off system and transport code. A disruption predictor has been derived based on MHD activity observations and statistical analysis. Sawtooth characteristics during ECRH are investigated and coupling between m = 1 and m/n = 2/1 modes is studied. Detachment features of HL-2A divertor are numerically and experimentally studied using the code SOLPS5.0 and measured data. The long divertor legs and thin divertor throats in HL-2A pose MHD shaping problems resulting in momentum losses even at low densities and strongly enhanced main chamber losses.

The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving β_T and β_N values of 17% and 5.7 (%m T MA⁻¹), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable β in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.

Substantial advances have been made on the Mega Ampère Spherical Tokamak (MAST). The parameter range of the MAST confinement database has been extended and it now also includes pellet-fuelled discharges. Good pellet retention has been observed in H-mode discharges without triggering an ELM or an H/L transition during peripheral ablation of low speed pellets. Co-ordinated studies on MAST and DIII-D demonstrate a strong link between the aspect ratio and the beta scaling of H-mode energy confinement, consistent with that obtained when MAST data were merged with a subset of the ITPA database. Electron and ion ITBs are readily formed and their evolution has been investigated. Electron and ion thermal diffusivities have been reduced to values close to the ion neoclassical level. Error field correction coils have been used to determine the locked mode threshold scaling which is comparable to that in conventional aspect ratio tokamaks. The impact of plasma rotation on sawteeth has been investigated and the results have been well-modelled using the MISHKA-F code. Alfvén cascades have been observed in discharges with reversed magnetic shear. Measurements during off-axis NBCD and heating are consistent with classical fast ion modelling and indicate efficient heating and significant driven current. Central electron Bernstein wave heating has been observed via the O–X–B mode conversion process in special magnetically compressed plasmas. Plasmas with low pedestal collisionality have been established and further insight has been gained into the characteristics of filamentary structures at the plasma edge. Complex behaviour of the divertor power loading during plasma disruptions has been revealed by high resolution infra-red measurements.

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 10²⁰ m⁻³, which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.

This paper presents an overview of experimental results and progress made in investigating the link between magnetic topology, electric fields and transport in the TJ-II stellarator. The smooth change from positive to negative electric field observed in the core region as the density is raised is correlated with global and local transport data. A statistical description of transport is emerging as a new way to describe the coupling between profiles, plasma flows and turbulence. TJ-II experiments show that the location of rational surfaces inside the plasma can, in some circumstances, provide a trigger for the development of core transitions, providing a critical test for the various models that have been proposed to explain the appearance of transport barriers in relation to magnetic topology. In the plasma core, perpendicular rotation is strongly coupled to plasma density, showing a reversal consistent with neoclassical expectations. In contrast, spontaneous sheared flows in the plasma edge appear to be coupled strongly to plasma turbulence, consistent with the expectation for turbulent driven flows. The local injection of hydrocarbons through a mobile limiter and the erosion produced by plasmas with well-known edge parameters opens the possibility of performing carbon transport studies, relevant for understanding co-deposit formation in fusion devices.

The inertial confinement fusion (ICF) programme, the high-average-power lasers (HAPL) programme, and the heavy ion fusion (HIF) programme are making long-term investments to establish the scientific and technical basis for an economically and environmentally attractive fusion power source. In the near term, the National Ignition Campaign is expected to establish the scientific and technical basis for ignition and gain on the National Ignition Facility. The results of these experiments and the implications for target design will be incorporated into the long-term efforts to develop a viable power-plant concept including target production, chamber dynamics and driver.

Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the People's Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

The present status of experiments on zonal flows in magnetic confinement experiments is examined. The innovative use of traditional and modern diagnostics has revealed unambiguously the existence of zonal flows, their spatio-temporal characteristics, their relationship to turbulence and their effects on confinement. In particular, a number of observations have been accumulated on the oscillatory branch of zonal flows, named geodesic acoustic modes, suggesting the necessity for theories to give their proper description. In addition to these basic properties of zonal flows, several new methods have elucidated the processes of zonal flow generation from turbulence. Further investigation of the relationship between zonal flows and confinement is strongly encouraged as cross-device activity including low temperature, toroidal and linear devices.

We present an overview on one issue of practical interest for burning plasmas, i.e. whether fast ions and charged fusion products are sufficiently well confined such that they transfer their energy and/or momentum to the thermal plasma without appreciable degradation due to collective modes. In the present work, we address this issue by analysing theoretically the dynamics of shear Alfvén waves collectively excited by energetic particles in tokamak plasmas. Both linear physics, such as spectral and stability properties, and key non-linear wave and particle dynamics are identified and considered. We also discuss the investigations of such processes via computer simulations as well as the importance of benchmarking with existing or future experimental observations.