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The sessions of experiments EX/D, EX/S and EX/W covering (D) Plasma–material interactions, Divertors, Limiters, SOL, (S) Stability, (W) Wave–plasma interactions, Current drive, Heating and Energetic particles are summarized in this paper. These general topics are divided into two categories, which are (1) ITER oriented and (2) ITER/DEMO oriented corresponding to the subject of each topic and considering the road map of fusion research. Topics in the ITER oriented category are strongly linked to the present direction of the fusion research for ITER, namely tokamak research, whereas issues in the ITER/DEMO oriented category are generally common issues for both tokamaks and helical devices.

The contributions to the 22nd IAEA Fusion Energy Conference (FEC) in the categories of Fusion Technology (FT), ITER Activities (IT) and Safety and Economic Studies (SE) are reviewed. In the FT category, 68 papers were submitted, along with 57 papers submitted through the ITER Organisation in the IT category. Finally two papers were submitted in the SE category. The assembled body of work gave a good overview of the worldwide effort in fusion technology and particularly the prolific activity surrounding the ITER Design Review and the major progress with the ITER technology programme.

This is a summary of the reports presented to the 22nd IAEA Fusion Energy Conference, Magnetic confinement theory and modelling section (Geneva, October 2008). Many of the papers are devoted to the investigation of transport processes, in particular to the toroidal momentum transport. Simulation by gyrokinetic codes has been improved in many countries, and the number of available codes reaches several tens. Numerical developments tend to follow the same trend as improvements in the computation power. The timescale for plasma simulations is now comparable to the ion–ion collision time. To improve the predictions for ITER, the near future advances are the combination of gyrokinetic and fluid codes. Reports on stellarators confirm that in these devices the neoclassical transport dominates, but the influence of turbulent transport can play a role in improved confinement regimes and in the resilience of pressure profiles. The resonant magnetic perturbations, mitigating the ELMs, could brake the plasma rotation, increasing the danger of disruption. The problems on the scrape-off layer and the divertor attract a large number of theoretical works that could lead to a better understanding of periphery plasma processes. ITER and reactor studies have been presented, and calculations confirm that ITER can achieve Q = 10 or larger. It has also been shown that the alpha-particle diffusion due to drift driven ITG turbulence will be relatively small in ITER, uncertainty remains in the magnitude of alpha-particle diffusion due to Alfvén waves.

Laser and ion beam inertial fusion research has made remarkable progress during the last two years. As a highlight of the 22nd IAEA Fusion Energy Conference among over 30 presentations the National Ignition Facility has been reported for completion as of 27 March 2009 to achieve the first fusion shot within 2010 or 2011 with a central ignition scheme. The LFEX and OMEGA-EP fast ignition lasers have also been completed to start fast ignition feasibility studies as early as in 2009. Relativistic physics revealed many new findings on hot electron energy transport and production. For example, several proposals have been reported to control the hot electron divergence angle. Laser technology has challenged to meet the specification required for fast ignition experiments such as in the LFEX and OMEGA-EP. Target development is making steady progress to be ready for the forthcoming fast ignition integral and ion beam experiments. The LIFE engine (Laser Inertial-Confinement Fusion Fission Energy) was announced from the USA to accelerate the energy production making use of the NIF type fusion outputs in order to prepare for 'beyond the ignition' as early as 2020. KOYO-F from Osaka University, Z-machine and HiPER European projects have been reported.

The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, 'bootstrap-aligned' equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch–Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears in the X-point region. Real time control techniques are currently being applied to EC multiple independent power supplies and beam launchers, e.g. to control the plasma current in fully non-inductive conditions, and the plasma elongation through current broadening by far-off-axis heating at constant shaping field.

Since the last IAEA conference, the scientific programme of JET has focused on the qualification of the integrated operating scenarios for ITER and on physics issues essential for the consolidation of design choices and the efficient exploitation of ITER. Particular attention has been given to the characterization of the edge plasma, pedestal energy and edge localized modes (ELMs), and their impact on plasma facing components (PFCs). Various ELM mitigation techniques have been assessed for all ITER operating scenarios using active methods such as resonant magnetic field perturbation, rapid variation of the radial field and pellet pacing. In particular, the amplitude and frequency of type I ELMs have been actively controlled over a wide parameter range (q₉₅ = 3–4.8, β_N ⩽ 3.0) by adjusting the amplitude of the n = 1 external perturbation field induced by error field correction coils. The study of disruption induced heat loads on PFCs has taken advantage of a new wide-angle viewing infrared system and a fast bolometer to provide a detailed account of time, localization and form of the energy deposition. Specific ITER-relevant studies have used the unique JET capability of varying the toroidal field (TF) ripple from its normal low value δ_BT = 0.08% up to δ_BT = 1% to study the effect of TF ripple on high confinement-mode plasmas. The results suggest that δ_BT < 0.5% is required on ITER to maintain adequate confinement to allow Q_DT = 10 at full field. Physics issues of direct relevance to ITER include heat and toroidal momentum transport, with experiments using power modulation to decouple power input and torque to achieve first experimental evidence of inward momentum pinch in JET and determine the threshold for ion temperature gradient driven modes. Within the longer term JET programme in support of ITER, activities aiming at the modification of the JET first wall and divertor and the upgrade of the neutral beam and plasma control systems are being conducted. The procurement of all components will be completed by 2009 with the shutdown for the installation of the beryllium wall and tungsten divertor extending from summer 2009 to summer 2010.

Recent JT-60U experimental results towards the establishment of advanced tokamak (AT) operation are reviewed. We focused on the further expansion of the operational regime of AT plasmas towards higher β_N regime with wall stabilization. After the installation of ferritic steel tiles in 2005, the high power heating in a large plasma cross-section in which the wall stabilization is expected has been possible. In 2007, the modification of power supply of NBIs improved the flexibility of the heating profile in long-pulse plasmas. The investigation of key physics issues for the establishment of steady-state AT operation is also in progress using new diagnostics and improved heating systems. In weak magnetic shear plasma, high β_N ∼ 3 exceeding the ideal MHD limit without a conducting wall ( ) is sustained for ∼5 s (∼3τ_R) with RWM stabilization by a toroidal rotation at the q = 2 surface. External current drivers of negative-ion based NB and lower-hybrid waves together with a large bootstrap current fraction (f_BS) of 0.5 can sustain the whole plasma current of 0.8 MA for 2 s (1.5τ_R). In reversed magnetic shear plasma, high β_N ∼ 2.7 (β_p ∼ 2.3) exceeding with q_min ∼ 2.4 (q₉₅ ∼ 5.3), H_H98(y,2) ∼ 1.7 and f_BS ∼ 0.9 is obtained with wall stabilization. These plasma parameters almost satisfy the requirement of ITER steady-state scenario. In long-pulse plasmas with positive magnetic shear, a high β_NH_H98(y,2) of 2.6 with β_N ∼ 2.6 and H_H98(y,2) ∼ 1 is sustained for 25 s, significantly longer than the current diffusion time (∼14τ_R) without neoclassical tearing modes (NTMs). A high G-factor, (a major of fusion gain), of 0.54 and a large f_BS > 0.43 are suitable for ITER hybrid operation scenario. Based on the plasma for ITER hybrid operation scenario, the high β_N of 2.1 with good thermal plasma confinement of H_H98(y,2) > 0.85 is sustained for longer than 12 s at and f_rad > 0.79. Physics studies for the development of AT plasmas, physics studies of H-mode, pedestal and ELM characteristics and physics studies on impurity transport, SOL/divertor plasmas and plasma–wall interactions are also in progress. The active NTM stabilization system using modulated ECCD, which is synchronized to rotating island, has been developed and the efficiency of modulated ECCD in m/n = 2/1 NTM stabilization has been demonstrated. The intrinsic toroidal rotation driven by the ion pressure gradient and by the ECH is confirmed. The dedicated H-mode and pedestal experiments indicate two scalings, H-factor evaluated for the core plasma as and pedestal width scaling of . New fast diagnostics with high spatial and temporal resolutions reveals the different structures of pedestal pressure between co- and counter-rotating plasma, resulting in different ELM sizes determined by the radial penetration depth of the ELM crash. The tungsten accumulation becomes more significant with increasing toroidal rotation in the counter-direction.

DIII-D research is providing key information for the design and operation of ITER. Investigations of axisymmetric stability and of edge-localized mode (ELM) suppression with resonant magnetic perturbations have helped provide the physics basis for new axisymmetric and non-axisymmetric control coils in ITER. Discharges that simulate ITER operating scenarios in conventional H-mode, advanced inductive, hybrid and steady state regimes have achieved normalized performance consistent with ITER's goals for fusion performance. Stationary discharges with high β_N and 90% non-inductive current that project to Q = 5 in ITER have been sustained for a current relaxation time (∼2.5 s), and high beta wall-stabilized discharges with fully non-inductive current drive have been sustained for more than one second. Detailed issues of plasma control have been addressed, including the development of a new large-bore startup scenario for ITER. DIII-D research also contributes to the basis for reliable operation in ITER, through active control of the chief performance-limiting instabilities. Simultaneous stabilization of neoclassical tearing modes (by localized current drive) and resistive wall modes (by magnetic feedback) has allowed stable operation at high beta and low rotation. In research aimed at improving the lifetime of material surfaces near the plasma, recent experiments have investigated several approaches to mitigation of disruptions, including injection of low-Z gas and low-Z pellets, and have shown the conditions that minimize core impurity accumulation during radiative divertor operation. Investigation of carbon erosion, transport and co-deposition with hydrogenic species, and methods for the removal of co-deposits, will contribute to the physics basis for initial operation of ITER with a carbon divertor. A broad research programme provides the physics basis for predicting the performance of ITER. Recent key results include the discovery that the L–H power threshold is reduced with low neutral beam torque, and the development of a successful model for prediction of the H-mode pedestal height in DIII-D. Research areas with the potential to improve ITER's performance include the demonstration of ELM-free 'quiescent H-mode' discharges with both co- and counter-neutral beam injection, and validation of the predicted torque generated by static, non-axisymmetric magnetic fields. New diagnostics provide detailed benchmarking of turbulent transport codes and direct measurements of the anomalous transport of fast ions by Alfvén instabilities. Successful comparison of experiment and modelling for off-axis neutral beam current drive provides the basis for more flexible current profile control in advanced scenarios.

ASDEX Upgrade was operated with a fully W-covered wall in 2007 and 2008. Stationary H-modes at the ITER target values and improved H-modes with H up to 1.2 were run without any boronization. The boundary conditions set by the full W wall (high enough ELM frequency, high enough central heating and low enough power density arriving at the target plates) require significant scenario development, but will apply to ITER as well. D retention has been reduced and stationary operation with saturated wall conditions has been found. Concerning confinement, impurity ion transport across the pedestal is neoclassical, explaining the strong inward pinch of high-Z impurities in between ELMs. In improved H-mode, the width of the temperature pedestal increases with heating power, consistent with a scaling. In the area of MHD instabilities, disruption mitigation experiments using massive Ne injection reach volume averaged values of the total electron density close to those required for runaway suppression in ITER. ECRH at the q = 2 surface was successfully applied to delay density limit disruptions. The characterization of fast particle losses due to MHD has shown the importance of different loss mechanisms for NTMs, TAEs and also beta-induced Alfven eigenmodes (BAEs). Specific studies addressing the first ITER operational phase show that O1 ECRH at the HFS assists reliable low-voltage breakdown. During ramp-up, additional heating can be used to vary l_i to fit within the ITER range. Confinement and power threshold in He are more favourable than in H, suggesting that He operation could allow us to assess H-mode operation in the non-nuclear phase of ITER operation.

The main results of the Tore Supra experimental programme in the years 2007–2008 are reported. They document significant progress achieved in the domain of steady-state tokamak research, as well as in more general issues relevant for ITER and for fusion physics research. Three areas are covered: ITER relevant technology developments and tests in a real machine environment, tokamak operational issues for high power and long pulses, and fusion plasma physics. Results presented in this paper include test and validation of a new, load-resilient concept of ion cycotron resonance heating antenna and of an inspection robot operated under ultra-high vacuum and high temperature conditions; an extensive experimental campaign (5 h of plasma) aiming at deuterium inventory and carbon migration studies; real-time control of sawteeth by electron cyclotron current drive in the presence of fast ion tails; ECRH-assisted plasma start-up studies; dimensionless scalings of transport and turbulence; transport experiments using active perturbation methods; resistive and fast-particle driven MHD studies. The potential role of Tore Supra in the worldwide fusion programme before the start of ITER operation is also discussed.

First divertor plasma configuration in Experimental Advanced Superconducting Tokamak (EAST) was obtained in the second campaign after the last IAEA meeting. To achieve long pulse diverted plasma discharges, new capabilities including the fully actively water cooled in-vessel components, current drive and heating systems, diagnostics and real-time plasma control algorithm were developed. Pre-programmed shape and feedback control of plasma position and current (RZIP) produced a variety of shaped plasma configurations, covering most of the configurations foreseen at the design stage of the machine. Control algorithm based on real-time equilibrium reconstruction and iso-flux control for the last closed magnetic flux surface (RTEFIT/ISOFLUX) has also been realized. A number of operational issues, such as plasma initiation and ramp up under constraints of superconducting coils were successfully investigated. First LHCD experiments demonstrated long pulse discharges longer than 20 s and nearly full non-inductive current drive. The physical engineering capability on the superconducting magnetic system was assessed by simulating discharges. Since the last IAEA meeting, experiments in HT-7 have been focusing on long pulse operation to support the EAST experiments on both physics and technical aspects. Long pulse discharges up to 400 s have now been achieved in HT-7. Investigation of sawtooth activities in ohmic and LHCD plasmas supports the turbulence model instead of the fast reconnection of the m = 1 magnetic island. Coexistence of electron mode and ion mode in high density ohmic plasmas has been observed by 2D ECE imaging (ECEI) in HT-7. The spectral characteristics of geodesic acoustic mode at the plasma boundary have been investigated by Langmuir probe arrays.

Significant experimental advances have been made on the HL-2A tokamak along with substantial improvement and development of the hardware. A spontaneous particle transport barrier has been observed in Ohmic discharges without any external momentum input. The barrier was evidenced by a density perturbation study using modulated supersonic molecular beam injection (SMBI) and microwave reflectometry. The new features of the non-local transport effect induced with SMBI have been analysed. The three-dimensional spectral structures of the low frequency zonal flow, the geodesic acoustic mode (GAM) and the quasi-mode-like low frequency fluctuations have been observed simultaneously for the first time. In addition, the spectral structure of the density fluctuations of GAM was also identified. The e-fishbone instability excited by energetic electrons deviated from Maxwellian distribution has been investigated via a 10-channel CdTe hard x-ray detector. It was found that the e-fishbone was correlated with the existence of energetic electrons of 30–70 keV. The MHD experiment has indicated that the suppression of m/n = 2/1 tearing modes may be sustained by ECRH modulation at a frequency of about 10 Hz.

Spontaneous increases in plasma density, up to ∼1.6 times the Greenwald value, are observed in FTU with lithized walls. These plasmas are characterized by profile peaking up to the highest obtained densities. The transport analysis of these discharges shows a 20% enhancement of the energy confinement time, with respect to the ITER97 L-mode scaling, correlated with a threshold in the peaking factor. It has been found that 0.4 MW of ECRH power, coupled at q = 2 surface, are sufficient to avoid disruptions in 0.5 MA discharges. Direct heating of magnetic islands produced by MHD modes determines current quench delay or avoidance. Supra-thermal electrons generated by 0.5 MW of lower hybrid power are sufficient to trigger precursors of the electron-fishbone instability. Evidence of spatial redistribution of fast electrons, on the ∼100 µs typical mode timescale, is shown by the fast electrons bremsstrahlung diagnostic. From the presence of new magnetic island induced accumulation points in the continuous spectrum of the shear Alfvén wave spectrum, the existence of new magnetic island induced Alfvén eigenmodes (MiAE) is suggested. Due to the frequency dependence on the magnetic island size, the feasibility of utilizing MiAE continuum effects as a novel magnetic island diagnostic is also discussed. Langmuir probes have been used on FTU to identify hypervelocity (10 km s⁻¹), micrometre size, dust grains. The Thomson scattering diagnostic was also used to characterize the dust grains, present in the FTU vacuum chamber, following a disruption. Analysis of the broad emitted light spectrum was carried out and a model taking into account the particle vaporization is compared with the data. A new oblique ECE diagnostic has been installed and the first results, both in the presence of lower hybrid or electron cyclotron waves, are being compared with code predictions. A time-of-flight refractometer at 60 GHz, which could be a good candidate for the ITER density feedback control system, has also been tested.

This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006–2008. Active flow drive, using mode converted ion cyclotron waves, has been observed for the first time in a tokamak plasma, using a mix of D and ³He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower hybrid range of frequencies (LHRF) microwaves have been used for efficient current drive, current profile modification and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage, has been developed and tested. During gas-puff mitigation of disruptions, induced MHD instabilities cause the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading and loss of runaway electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer E_r profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q₉₅ ⩽ 3 with ion drift away from the active X-point, shows very good energy confinement with a strong temperature pedestal, a weak density pedestal, and no evidence of particle or impurity accumulation, without the need for ELMs or any additional edge density regulation mechanism.

Remarkable progress in the physical parameters of net-current free plasmas has been made in the Large Helical Device (LHD) since the last Fusion Energy Conference in Chengdu, 2006 (Motojima et al2007 Nucl. Fusion47 S668). The beta value reached 5% and a high-beta state beyond 4.5% from the diamagnetic measurement has been maintained for longer than 100 times the energy confinement time. The density and temperature regimes have also been extended. The central density has exceeded 1 × 10²¹ m⁻³ due to the formation of an internal diffusion barrier. The ion temperature has reached 5.2 keV at the density of 1.6 × 10¹⁹ m⁻³, which is associated with the suppression of ion heat conduction loss. Although these parameters have been obtained in separated discharges, each fusion-reactor relevant parameter has elucidated the potential of net-current free heliotron plasmas. Diversified studies in recent LHD experiments are reviewed in this paper.

The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l_i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β_N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f_NI ∼ 71%. Instabilities driven by super-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.

Several improvements to the MAST plant and diagnostics have facilitated new studies advancing the physics basis for ITER and DEMO, as well as for future spherical tokamaks (STs). Using the increased heating capabilities P_NBI ⩽ 3.8 MW H-mode at I_p = 1.2 MA was accessed showing that the energy confinement on MAST scales more weakly with I_p and more strongly with B_t than in the ITER IPB98(y, 2) scaling. Measurements of the fuel retention of shallow pellets extrapolate to an ITER particle throughput of 70% of its original designed total throughput capacity. The anomalous momentum diffusion, χ_ϕ, is linked to the ion diffusion, χ_i, with a Prandtl number close to P_ϕ ≈ χ_ϕ/χ_i ≈ 1, although χ_i approaches neoclassical values. New high spatial resolution measurements of the edge radial electric field, E_r, show that the position of steepest gradients in electron pressure and E_r (i.e. shearing rate) are coincident, but their magnitudes are not linked. The T_e pedestal width on MAST scales with rather than ρ_pol. The edge localized mode (ELM) frequency for type-IV ELMs, new in MAST, was almost doubled using n = 2 resonant magnetic perturbations from a set of four external coils (n = 1, 2). A new internal 12 coil set (n ⩽ 3) has been commissioned. The filaments in the inter-ELM and L-mode phase are different from ELM filaments, and the characteristics in L-mode agree well with turbulence calculations. A variety of fast particle driven instabilities were studied from 10 kHz saturated fishbone like activity up to 3.8 MHz compressional Alfvén eigenmodes. Fast particle instabilities also affect the off-axis NBI current drive, leading to fast ion diffusion of the order of 0.5 m² s⁻¹ and a reduction in the driven current fraction from 40% to 30%. EBW current drive start-up is demonstrated for the first time in a ST generating plasma currents up to 55 kA. Many of these studies contributed to the physics basis of a planned upgrade to MAST.

This paper presents the latest results on confinement studies in the TJ-II stellarator. The inherently strong plasma–wall interaction of TJ-II has been successfully reduced after lithium coating by vacuum evaporation. Besides H retention and low Z, Li was chosen because there exists a reactor-oriented interest in this element, thus giving special relevance to the investigation of its properties. The Li-coating has led to important changes in plasma performance. Particularly, the effective density limit in NBI plasmas has been extended reaching central values of 8 × 10¹⁹ m⁻³ and T_e ≈ 250–300 eV, with peaked density, rather flat T_e profiles and higher ion temperatures. Due to the achieved density control, a second type of transition has been added to the low density ones previously observed in ECRH plasmas: higher density transitions characterized by the fall in Hα emission, the onset of steep density gradient and the reduction in the turbulence; which are characteristics of transition to the H mode. Confinement studies in ECH plasmas indicate that lowest order magnetic resonances, even in a low shear environment, locally reduce the effective electron heat diffusivities, while Alfven eigenmodes destabilized in NBI plasmas can influence fast ion confinement.

With the exploration of the MA plasma current regime in up to 0.5 s long discharges, RFX-mod has opened new and very promising perspectives for the reversed field pinch (RFP) magnetic configuration, and has made significant progress in understanding and improving confinement and in controlling plasma stability. A big leap with respect to previous knowledge and expectations on RFP physics and performance has been made by RFX-mod since the last 2006 IAEA Fusion Energy Conference. A new self-organized helical equilibrium has been experimentally achieved (the Single Helical Axis—SHAx—state), which is the preferred state at high current. Strong core electron transport barriers characterize this regime, with electron temperature gradients comparable to those achieved in tokamaks, and by a factor of 4 improvement in confinement time with respect to the standard RFP. RFX-mod is also providing leading edge results on real-time feedback control of MHD instabilities, of general interest for the fusion community.

We have increased substantially the electron and ion temperatures, the electron density, and the total beta in plasmas with improved energy confinement in the Madison Symmetric Torus (MST). The improved confinement is achieved with a well-established current profile control technique for reduction of magnetic tearing and reconnection. A sustained ion temperature >1 keV is achieved with intensified reconnection-based ion heating followed immediately by current profile control. In the same plasmas, the electron temperature reaches 2 keV, and the electron thermal diffusivity drops to about 2 m² s⁻¹. The global energy confinement time is 12 ms. This and the reported temperatures are the largest values yet achieved in the reversed-field pinch (RFP). These results were attained at a density ∼10¹⁹ m⁻³. By combining pellet injection with current profile control, the density has been quadrupled, and total beta has nearly doubled to a record value of about 26%. The Mercier criterion is exceeded in the plasma core, and both pressure-driven interchange and pressure-driven tearing modes are calculated to be linearly unstable, yet energy confinement is still improved. Transient momentum injection with biased probes reveals that global momentum transport is reduced with current profile control. Magnetic reconnection events drive rapid momentum transport related to large Maxwell and Reynolds stresses. Ion heating during reconnection events occurs globally, locally, or not at all, depending on which tearing modes are involved in the reconnection. To potentially augment inductive current profile control, we are conducting initial tests of current drive with lower-hybrid and electron-Bernstein waves.

Experiments and simulations to achieve high values of plasma parameters at the Globus-M spherical tokamak (ST) at moderate absolute auxiliary heating power (up to 0.8 MW) and high specific heating power (up to 2–3 MW m⁻³) are described. Important distinguishing features are the low edge safety factor range, which is unusual for STs, 2.7 < q < 5 and small plasma–outer wall space (3–5 cm). High ion heating efficiency with neutral beam injection (NBI) was demonstrated. Results of numerical simulation of fast ion trajectories are described and fast ion generation during the NBI and ion cyclotron resonance heating is discussed. Investigations on their confinement and slowing down are also presented. Reasons for achievement of high IC heating efficiency are outlined. Reliable H-mode regime achievement is described. Transport ASTRA modelling demonstrated that during NB heated H-mode ion heat diffusivity remains neoclassical and the particle diffusion coefficient inside transport barrier decreases significantly. Analysis was performed of divertor tile and special probe surfaces after irradiation by plasma during a large number of shots (3000–10 000 shots). Mixed layer composition is measured and deuterium retention in different tokamak first wall areas is estimated. Plasma jet injection experiments with upgraded plasma jet are described. Jet penetration to the plasma centre with immediate increase of density and temperature drop is proved and analogy with pellet injection is outlined.

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high-energy-density (HED) science, is nearing completion at Lawrence Livermore National Laboratory (LLNL). NIF, a 192-beam Nd-glass laser facility, will produce 1.8 MJ, 500 TW of light at the third-harmonic, ultraviolet light of 351 nm. The NIF project is scheduled for completion in March 2009. Currently, all 192 beams have been operationally qualified and have produced over 4.0 MJ of light at the fundamental wavelength of 1053 nm, making NIF the world's first megajoule laser. The principal goal of NIF is to achieve ignition of a deuterium–tritium (DT) fuel capsule and provide access to HED physics regimes needed for experiments related to national security, fusion energy and for broader scientific applications.

The plan is to begin 96-beam symmetric indirect-drive ICF experiments early in FY2009. These first experiments represent the next phase of the National Ignition Campaign (NIC). This national effort to achieve fusion ignition is coordinated through a detailed plan that includes the science, technology and equipment such as diagnostics, cryogenic target manipulator and user optics required for ignition experiments. Participants in this effort include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory and the University of Rochester Laboratory for Energetics (LLE). The primary goal for NIC is to have all of the equipment operational and integrated into the facility soon after project completion and to conduct a credible ignition campaign in 2010. When the NIF is complete, the long-sought goal of achieving self-sustaining nuclear fusion and energy gain in the laboratory will be much closer to realization.

Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of inertial fusion energy (IFE) and will likely focus the world's attention on the possibility of an ICF energy option. NIF experiments to demonstrate ignition and gain will use central-hot-spot (CHS) ignition, where a spherical fuel capsule is simultaneously compressed and ignited. The scientific basis for CHS has been intensively developed (Lindl 1998 Inertial Confinement Fusion: the Quest for Ignition and Energy Gain Using Indirect Drive (New York: American Institute of Physics)) and has a high probability of success. Achieving ignition with CHS will open the door for other advanced concepts, such as the use of high-yield pulses of visible wavelength rather than ultraviolet and fast ignition concepts (Tabak et al 1994 Phys. Plasmas1 1626–34, Tabak et al 2005 Phys. Plasmas12 057305). Moreover, NIF will have important scientific applications in such diverse fields as astrophysics, nuclear physics and materials science.

This paper summarizes the design, performance and status of NIF, experimental plans for NIC, and will present laser inertial confinement fusion–fission energy (LIFE) as a path to achieve carbon-free sustainable energy.

A number of experiments have been undertaken at the Rutherford Appleton Laboratory that were designed to investigate the physics of fast electron transport relevant to fast ignition inertial fusion. The laser, operating at a wavelength of 1054 nm, provided pulses of up to 350 J of energy on target in a duration that varied in the range 0.5–5 ps and a focused intensity of up to 10²¹ W cm⁻². A dependence of the divergence of the fast electron beam with intensity on target has been identified for the first time. This dependence is reproduced in two-dimensional particle-in-cell simulations and has been found to be an intrinsic property of the laser–plasma interaction. A number of ideas to control the divergence of the fast electron beam are described. The fractional energy transfer to the fast electron beam has been obtained from calibrated, time-resolved, target rear-surface radiation temperature measurements. It is in the range 15–30%, increasing with incident laser energy on target. The fast electron temperature has been measured to be lower than the ponderomotive potential energy and is well described by Haines' relativistic absorption model.

Since the approval of the first phase of the Fast-Ignition Realization Experiment (FIREX-I), we have devoted our efforts to designing advanced targets and constructing a petawatt laser, which will be the most energetic petawatt laser in the world. Scientific and technological improvements are required to efficiently heat the core plasma. There are two methods that can be used to enhance the coupling efficiency of the heating laser to the thermal energy of the compressed core plasma: adding a low-Z foam layer to the inner surface of the cone and employing a double cone. The implosion performance can be improved in three ways: adding a low-Z plastic layer to the outer surface of the cone, using a Br-doped plastic ablator and evacuating the target centre. An advanced target for FIREX-I was introduced to suit these requirements. A new heating laser (LFEX) has been constructed that is capable of delivering an energy of 10 kJ in 10 ps with a 1 ps rise time. A fully integrated fast-ignition experiment is scheduled for 2009.

Magnetic islands are a ubiquitous feature of magnetically confined plasmas. They arise as the result of plasma instabilities as well as externally imposed symmetry-breaking perturbations. In the core, effective suppression techniques have been developed. Even thin islands, however, are observed to have nonlocal effects on the profiles of rotation and current. This has stimulated interest in using magnetic islands to control plasma transport, particularly in the edge. They are also of interest as a tool to improve our understanding of microscopic plasma dynamics.

This paper presents an overview of the results obtained during the Joint Experiments organized in the framework of the IAEA Coordinated Research Project on 'Joint Research Using Small Tokamaks' that have been carried out on the tokamaks CASTOR at IPP Prague, Czech Republic (2005), T-10 at RRC 'Kurchatov Institute', Moscow, Russia (2006), and the most recent one at ISTTOK at IST, Lisbon, Portugal, in 2007. Experimental programmes were aimed at diagnosing and characterizing the core and the edge plasma turbulence in a tokamak in order to investigate correlations between the occurrence of transport barriers, improved confinement, electric fields and electrostatic turbulence using advanced diagnostics with high spatial and temporal resolution. On CASTOR and ISTTOK, electric fields were generated by biasing an electrode inserted into the edge plasma and an improvement of the global particle confinement induced by the electrode positive biasing has been observed. Geodesic acoustic modes were studied using heavy ion beam diagnostics on T-10 and ISTTOK and correlation reflectometry on T-10. ISTTOK is equipped with a gallium jet injector and the technical feasibility of gallium jets interacting with plasmas has been investigated in pulsed and ac operation. The first Joint Experiments have clearly demonstrated that small tokamaks are suitable for broad international cooperation to conduct dedicated joint research programmes. Other activities within the IAEA Coordinated Research Project on Joint Research Using Small Tokamaks are also overviewed.