Table of contents

The group of 25 articles published in this special issue of Nuclear Fusion aims to monitor the progress made with experiments on fusion physics that have been conducted worldwide up to the end of 2004. These articles are based on overview reports from the various experimental teams presented at the Fusion Energy Conference (FEC 2004). This conference was organized by the IAEA together with the Portuguese host organization CFN-IST and was held in Vilamoura, Portugal, in early November 2004. The overviews presented at the conference have been rewritten and extended for the purpose of this special issue and submitted to the standard double-referee peer-review of Nuclear Fusion. Most teams have made use of this opportunity. Therefore this issue, which also includes four conference summaries, presents a reasonably complete picture of the progress made since FEC 2002 in Lyon.

The articles are placed in the following sequence:

• Conference summaries

  • Theory of magnetic confinement

  • Experimental confinement, plasma-material interactions and innovative concepts

  • Experiments on stability, energetic particles, waves and current drive

  • Inertial confinement fusion

• Tokamaks

  • Performance: JT-60U, JET, DIII-D, ASDEX-U, C-MOD

  • Steady state/long pulse operation: Tore Supra, HT-7, TRIAM

  • Spherical tokamaks: MAST, NSTX

  • Tritium experiments: JET

  • Diagnostics and heating methods: JET (diagnostics), T-10 (ECRH and diagnostics) and FTU (LHH + ECRH)

  • New devices: HL-2A

  • Small devices

• Alternative magnetic confinement concepts

  • Stellarators: LHD, TJ-II

  • Reversed field pinches: MST

• Inertial confinement

  • Direct drive

  • Heavy ion beam fusion

Readers will also notice the supplementary issue of the journal (volume 45, issue 10A). This extra issue contains the 15-year overview report on progress in fusion research as written by the International Fusion Research Council (IFRC) under the editorial responsibility of the IFRC. Both issues together will give the interested reader a state-of-the-art picture of the progress in nuclear fusion research.

This article summarizes the papers on the theory of magnetic confinement presented at the 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1–6 November 2004.

This paper summarizes the results presented at the 20th IAEA Fusion Energy Conference 2004 in the sessions on confinement, plasma–material interactions and innovative confinement concepts. The highlights of the presentations are as follows. Long pulse operation has been achieved with high beta and high bootstrap fraction for periods much longer than the current diffusion time. The discharge scenario optimization and its extrapolation towards ITER have progressed remarkably. Significant progress has been made in the understanding of global confinement and transport physics. Progress has also been made in material and retention/migration studies. These results will greatly contribute to ITER and the ultimate goal of fusion power generation.

This is the summary paper for the subjects of plasma stability, energetic particles, waves and current drive at the 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1–6 November 2004. Material summarized herein was drawn from 65 contributed papers and 21 overview papers. The subject-wise distribution of contributed papers is shown in the following table:

Subject	Papers
Resistive wall modes	7
Disruptions	6
Neoclassical tearing modes	6
Edge localized modes and the edge pedestal	13
Other stability topics	9
Alfvén modes	9
Wave physics	10
Current drive	5
Total	65

Significant advances have been reported on the principal instabilities in magnetically confined plasmas, even extending to the burning plasma state. Wave-plasma physics is maturing and novel methods of current drive and non-inductive current generation are being developed.

Based on the results presented at the 20th IAEA Fusion Energy Conference 2004, this paper highlights the most important recent advances in inertial confinement fusion (ICF). With the construction of the National Ignition Facility (NIF) and the Laser Mégajoule facility and many improvements in the target design, the conventional indirect-drive approach is advancing steadily towards the demonstration of ignition and high gain. The development of the polar direct-drive concept also made the prospects for direct-drive ignition on the NIF very favourable. Substantial progress was reported on the exploration of the fast-ignition approach to ICF. Parallel to that, multi-wire Z-pinches have become a competitive driver option for achieving ignition at the lowest possible cost. In heavy-ion fusion, experiments have been devoted so far to studying the generation, transport, and final focusing of high-current ion beams. A new concept for a power plant with a heavy-ion driver, based on a cylindrical direct-drive target compressed and ignited (in the fast-ignition mode) by two separate beams of very energetic (E_i≳ 0.5 GeV u⁻¹) heavy ions, has been proposed.

Recent experimental results from steady-state advanced tokamak (AT) research on JT-60U are presented with emphasis on time scales longer in comparison with the characteristic time scales in plasmas. To achieve this, modification of the controls for the operation, heating and diagnostics systems have been carried out. As a result, ∼60 s current flat top and a ∼30 s H-mode are obtained. The long pulse modification has opened a door into a new domain for JT-60U. High normalized beta (β_N) of 2.3 is maintained for 22.3 s and 2.5 for 16.5 s in a high-β_p H-mode plasma. A standard ELMy H-mode plasma has also been extended and changes in the wall recycling on the longer time scale have been unveiled. The development and investigation of plasmas relevant to AT operation have been continued in long discharges as well as in discharges where higher NB power is available (⩽10 s). Higher β_N (∼3) is maintained for 6.2 s in a high-β_p H-mode plasma. High bootstrap current fraction (f_BS) of ∼75% is sustained for 7.4 s in a reversed shear plasma. Neo-classical tearing mode (NTM) suppression by localized ECCD is found to be more effective with ECRF injection preceding the mode saturation. The mode is suppressed with less power compared to the injection after the mode saturates. The domain of the NTM suppression experiments is extended to the high-β_N regime, and the effectiveness of m/n = 3/2 mode suppression by ECCD is demonstrated at β_N ∼ 2.5–3. Genuine tokamak plasma start up without a central solenoid is demonstrated. In a current hole region, it is shown that no scheme drives current in any direction. Detailed measurement of energetic ions in both space and energy showed dynamic change in the energetic ion profile due to collective instabilities. The impact of toroidal plasma rotation on ELM behaviour is clarified in the grassy ELM and QH domains. Retention of hydrogen isotopes in the divertor tiles is analysed.

High density and high confinement operation in ELMy H-mode is confirmed at or above the normalized parameters foreseen for the ITER operating point (H_98(y,2) ∼ 1, n/n_GW ∼ 1, β_N > 1.8 at q₉₅ ∼ 3). The scaling of the ELMy H-mode with β_N could be more favourable than that predicted by the IPB98(y,2) scaling. In ELMy H-mode, ion cyclotron current drive (ICCD) control of large sawteeth stabilized by fast particle has been demonstrated and the underlying neo-classical tearing modes (NTMs) and sawtooth physics is being refined. At high-density, Type I ELMy H-modes show trends that would lead to marginally acceptable ELMs on ITER. Type II ELM regime has been produced, though under very restrictive conditions. Type III ELMy operation with radiation fractions up to 95% has been demonstrated by seeding of N₂ in H-modes and could extrapolate to Q = 10 ITER operation, albeit at high current (17 MA). The mitigation of Type I ELMs, nevertheless, remains a challenge. Considerable progress has been obtained in internal transport barrier (ITB) plasmas, with operation at central densities close to the Greenwald density or/and low toroidal rotation or/and high triangularity. Demonstrations of full current drive and successful simultaneous real time control of safety factor and temperature profiles have been achieved in ITB plasmas. Physics of resistive wall modes (RWMs) has been compared with theory, showing favourable scaling for ITER. High β_N ∼ 2.8 operation of hybrid modes (also called improved H-modes) has been obtained with dominant neutral beam heating. Hybrid modes with dominant ion cyclotron resonance heating (ICRH) have also been achieved. Trace tritium experiments yielded valuable information on particle transport in H-mode, ITB and hybrid regimes. In Type I ELMy plasmas, successful tests of the conjugate-T ICRH scheme have been achieved as well as lower hybrid coupling at ITER-relevant 10–11 cm distances. Reduced D and T fuel retention has been observed, which could relate to operation with vertical targets in the divertor and/or lower (ITER-like) vessel temperature. It is confirmed that erosion occurs predominantly on the main chamber surfaces, with possible benefits for T retention in ITER, although consequences for the metallic first wall lifetime need to be assessed. Disruption and ELM studies indicate that transient power deposition could be less constraining than expected for the ITER divertor, but more challenging for the metallic first wall. Alpha particle tomography and direct observation of alpha particle slowing down have been made possible by γ-spectroscopy. Measurements of Alfvén cascades have been improved by a new interferometric technique. Promising tests of ITER relevant neutron counting detectors have been conducted.

Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance and basic tokamak physics has been made by the DIII-D team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than the value specified for Q = 10 in ITER reference scenario, under stationary conditions. Discharges have also been demonstrated in DIII-D with enhanced performance under stationary conditions that project to Q ∼ 10 for longer than 1 h in ITER at reduced current, if such a mode of operation can be realized in ITER. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, edge localized mode avoidance and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behaviour in the electron heat transport, rotation in plasmas in the absence of external torque, measurements in the edge pedestal region and plasma fuelling. Understanding these phenomena at a fundamental level contributes to development and ultimately the optimization of tokamak scenarios.

Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν^*, n/n_GW and T_e/T_i, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

Research on the Alcator C-Mod tokamak has emphasized RF heating, self-generated flows, momentum transport, scrape-off layer (SOL) turbulence and transport and the physics of transport barrier transitions, stability and control. The machine operates with P_RF up to 6 MW corresponding to power densities on the antenna of 10 MW m⁻². Analysis of rotation profile evolution, produced in the absence of external drive, allows transport of angular momentum in the plasma core to be computed and compared between various operating regimes. Momentum is clearly seen diffusing and convecting from the plasma edge on time scales similar to the energy confinement time and much faster than neo-classical transport. SOL turbulence and transport have been studied with fast scanning electrostatic probes situated at several poloidal locations and with gas puff imaging. Strong poloidal asymmetries are found in profiles and fluctuations, confirming the essential ballooning character of the turbulence and transport. Plasma topology has a dominant effect on the magnitude and direction of both core rotation and SOL flows. The correlation of self-generated plasma flows and topology has led to a novel explanation for the dependence of the H-mode power threshold on the ∇B drift direction. Research into internal transport barriers has focused on control of the barrier strength and location. The foot of the barrier could be moved to larger minor radius by lowering q or B_T. The barriers, which are produced in C-Mod by off-axis RF heating, can be weakened by the application of on-axis power. Gyro-kinetic simulations suggest that the control mechanism is due to the temperature dependence of trapped electron modes which are destabilized by the large density gradients. A set of non-axisymmetric coils was installed allowing intrinsic error fields to be measured and compensated. These also enabled the determination of the mode locking threshold and, by comparison with data from other machines, provided the first direct measurement of size scaling for the threshold. The installation of a new inboard limiter resulted in the reduction of halo currents following disruptions. This effect can be understood in terms of the change in plasma contact with the altered geometry during vertical displacement of the plasma column. Unstable Alfven eigenmodes (AE) were observed in low-density, high-power ICRF heated plasmas. The damping rate of stable AEs was investigated with a pair of active MHD antennae.

Key technological and physics issues related to long pulse operation required for a reactor are now being addressed in Tore Supra. This paper focuses on the experimental results from conditions where all the plasma facing components (PFCs) are actively cooled. Full steady-state plasma quantities as well as stable PFC temperature during pulses exceeding 6 min are reported. Particle balance and D retention in the steady-state are analysed. Physics phenomena related to continuous operation are observed in non-inductively driven plasmas, namely (i) a regime characterized by large sinusoidal oscillations of the central electron temperature, governed by a predator–prey mechanism of non-linear coupling between heat transport and plasma current; (ii) a strong synergy effect between lower hybrid and electron cyclotron waves; (iii) a peaked density profile generated by turbulent inward pinch. Integrated modelling activity at Tore Supra is also reported.

An overview of the progress with experiments in the HT-7 during 2003–4 is presented. The H-mode, negative reversed shear and high l_i operational scenarios were investigated for quasi-steady-state high performance plasma discharges. Ion Bernstein wave (IBW) heating at 30 MHz produced a typical edge H-mode plasma. Transport in both electron and ion channels were reduced in the outer half portion of the plasma. H-mode was produced in off-axis lower hybrid current drive (LHCD) plasmas by launching a lower hybrid wave (LHW) with a greater parallel refractive index, , of 3.1. The improved confinement status in such plasma discharges could be sustained for more than 400τ_E. A high inductance, l_i, plasma was created using a fast plasma current ramp down and sustained by central LHCD and IBW heating for a duration of >1 s with a strongly peaked electron temperature profile. The highest central electron temperature obtained was 4.5 keV. An increase in the energy confinement time with l_i was observed. It was found that the IBW heating could improve the plasma confinement at the same l_i if part of the LHW power was replaced by IBW. Stationary internal transport barriers with normalized performance H₈₉β_N > 1–3 were obtained with combined injection of LHW and IBW for a duration of several hundred energy confinement times in the weak negative magnetic shear (RS). Increasing the total injection power to 1 MW did not degrade the plasma confinement significantly in the RS operational scenario. Long pulse discharges were performed, using three feedback controls for the plasma current and position, the central line-averaged electron density and the magnetic swing flux of the transformer. The longest plasma discharge, with a duration of 240 s, T_e(0) ∼ 1 keV and the central electron density > 0.8 × 10¹⁹ m⁻³ was achieved in 2004. The wall saturation and refreshment (wall pumping) were first observed in such long pulse discharges. A fully LHW current driven plasma without using ohmic current in the central solenoid coils was sustained for 80 s. The main limitation for the pulse length was due to the recycling, which caused an uncontrollable rise in the electron density.

Experiments aiming at 'day long operation at high performance' have been carried out in TRIAM-1M. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation is studied under the localized plasma wall interaction conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. The formation and sustainment of an internal transport barrier (ITB) in enhanced current drive mode has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied.

Significant progress has been made on the Mega Ampere Spherical Tokamak (MAST) towards a fundamental understanding of transport, stability and edge physics and addressing technological issues for future large devices. Collaborative studies of the L–H transition with NSTX and ASDEX Upgrade confirm that operation in a connected double-null configuration significantly reduces the threshold power, P_thr. The MAST data provide support for a theory for the transition based on finite β drift wave turbulence suppression by self-generated zonal flows. Analysis of low and high field side density gradients in the H-mode pedestal provides support for an analytical model of the density pedestal width dependent on the neutral penetration depth. Adding MAST data to international confinement databases has enhanced confidence in scalings for ITER by significantly expanding the range of β and ε explored and indicates a slightly stronger ε dependence than in current scalings. Studies of core transport have been conducted for well-diagnosed L-mode, H-mode and internal transport barrier (ITB) discharges using TRANSP, and microstability and turbulence studies have been carried out using GS2. Linear micro-stability analysis indicates that ITG modes are typically unstable on all flux surfaces with growth rates that are comparable to the equilibrium E × B flow shearing rate. Mixing length estimates of transport coefficients from ITG (neglecting flow shear) give diffusion coefficients that are broadly comparable with observed thermal diffusivities. Non-linear, collisionless ETG calculations have been performed and suggest radially extended electrostatic streamers up to 100ρ_e across in radius. Transport from ITG could easily be suppressed in regions where the E × B shear flow rate, ω_SE, exceeds the ITG growth rate, possibly contributing to ITBs. Toroidal rotation, driven by neutral beam torque, is the dominant contribution to ω_SE via the v_ϕB_θ term in the radial electric field. Early edge localized mode activity on MAST is associated with the formation of narrow filamentary structures following field lines in the edge. These filaments rotate toroidally with the edge plasma and, away from the X-points, accelerate radially outwards from the edge up to 20 cm. Studies of disruptions on MAST demonstrate a complex evolution of core energy loss and resultant divertor power loads, including phases where the target heat flux width is broadened by a factor of 8. Observations of energetic particle modes driven by super-Alfvénic beam ions provide support for a model for the non-linear evolution of toroidal Alfvén eigenmodes (AEs) forming Bernstein–Green–Krushal waves. The AE activity reduces to low levels with increasing β. Plasma start-up without a central solenoid and in a manner compatible with future large spherical tokamak (ST) devices has been demonstrated using breakdown at a quadrupole magnetic null. Closed flux surface plasmas with peak plasma currents up to 370 kA have been generated and sustained for 0.3 s. New error field correction coils have extended the operational space for low density plasmas and enabled scaling studies of error field induced locked mode formation in the ST.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high β_T in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high β_T (up to ∼40%) lasting for many energy confinement times. β_T can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a B_T dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfvén eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (n_T/n_D < 3%). Thermal tritium particle transport coefficients (D_T, v_T) are found to exceed neo-classical values in all regimes, except in ELMy H-modes at high densities and in the region of internal transport barriers (ITBs) in reversed shear plasmas. In ELMy H-mode dimensionless parameter scans, at q₉₅ ∼ 2.8 and triangularity δ = 0.2, the T particle transport scales in a gyro-Bohm manner in the inner plasma (r/a < 0.4), whilst the outer plasma particle transport scaling is more Bohm-like. Dimensionless parameter scans show contrasting behaviour for the trace particle confinement (increases with collisionality, ν^* and β) and bulk energy confinement (decreases with ν^* and is independent of β). In an extended ELMy H-mode data set, with ρ^*, ν^*, β and q varied but with neo-classical tearing modes (NTMs) either absent or limited to weak, benign core modes (4/3 or above), the multiparameter fit to the normalized diffusion coefficient in the outer plasma (0.65 < r/a < 0.8) gives D_T/B_ϕ ∼ ρ^*2.46ν^*−0.23β^−1.01q^2.03. In hybrid scenarios (q_min ∼ 1, low positive shear, no sawteeth), the T particle confinement is found to scale with increasing triangularity and plasma current. Comparing regimes (ELMy H-mode, ITB plasma and hybrid scenarios) in the outer plasma region, a correlation of high values of D_T with high values of v_T is seen. The normalized diffusion coefficients for the hybrid and ITB scenarios do not fit the scaling derived for ELMy H-modes. The normalized tritium diffusion scales with normalized poloidal Larmor radius in a manner close to gyro-Bohm , with an added inverse β dependence. The effects of ELMs, sawteeth and NTMs on the T particle transport are described. Fast-ion confinement in current-hole (CH) plasmas was tested in TTE by tritium neutral beam injection into JET CH plasmas. γ-rays from the reactions of fusion alpha and beryllium impurities (⁹Be(α, nγ)¹²C) characterized the fast fusion-alpha population evolution. The γ-decay times are consistent with classical alpha plus parent fast triton slowing down times (τ_Ts + τ_αs) for high plasma currents (I_p > 2 MA) and monotonic q-profiles. In CH discharges the γ-ray emission decay times are much lower than classical (τ_Ts+τ_αs), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker–Planck numerical code and modelled using TRANSP.

Some recent JET campaigns, with the introduction of a trace amount (n_T/n_D < 5%) of tritium into D plasmas and third harmonic ICRH acceleration of ⁴He, provided unique opportunities to test 'burning plasma' diagnostics. In particular, new approaches and techniques were investigated for the detection of neutrons, α particles and the fuel mixture. With regard to neutron detection, the recent activity covered aspects such as calibration and cross validation of the diagnostics, measurement of the spatial distribution of the neutrons, particle transport and neutron spectrometry. The first tests of some new neutron detection technologies were also undertaken during the Trace Tritium Experiment campaign. To improve JET's diagnostic capability in the field of α particles, a significant development programme was devoted to the measurement of their confinement and imaging with γ-ray spectroscopy. A new approach for the fusion community to measuring the fast ion losses, based on the activation technique, was also attempted for the first time on JET. An assessment of the neutral particle analyser's potential to determine the fuel mixture and the particle transport coefficients is under way.

This report summarizes the results of experimental turbulence investigations carried out at T-10 for more than 10 years. The turbulence characteristics were investigated using correlation reflectometry, multipin Langmuir probe (MLP) and heavy ion beam probe diagnostics. The reflectometry capabilities were analysed using 2D full-wave simulations and verified by direct comparison using a MLP. The ohmic and electron cyclotron resonance heated discharges show the distinct transition from the core turbulence, having complex spectral structure, to the unstructured one in the scrape-off layer. The core turbulence includes 'broad band, quasi-coherent' features, arising due to the excitation of rational surfaces with high poloidal m-numbers, with a low frequency near zero and specific oscillations at 15–30 kHz. All experimentally measured properties of low frequency and high frequency quasi-coherent oscillations are in good agreement with predictions of linear theory for the ion temperature gradient/dissipative trapped electron mode instabilities. Significant local changes in the turbulence characteristics were observed at the edge velocity shear layer and in the core near q = 1 radius after switching off the electron cyclotron resonance heating (ECRH). The local decrease in the electron heat conductivity and decrease in the turbulence level could be evidence of the formation of an electron internal transport barrier. The dynamic behaviour of the core turbulence was also investigated for the case of fast edge cooling and the beginning phase of ECRH.

An overview of the FTU results during the period 2003–4 is presented. A prototype ITER-relevant lower hybrid current drive (LHCD) launcher, the passive active multijunction, has been successfully tested (f = 8 GHz) showing high power handling and good coupling properties and current drive comparable to those of a conventional launcher. Effective electron and ion heating (via collisions) have been achieved with the 140 GHz ECRH systems up to 1.5 MW, as well as current drive (25 kA at n_e0 = 8 × 10¹⁹ m⁻³). The mitigation of disruptions has been studied with on-axis ECRH. Ion Bernstein wave studies have shown the importance of recycling in achieving improved confinement plasmas. Advanced tokamak scenarios are presented including repetitive pellet enhanced plasmas and electron internal transport barriers (e-ITBs). Very peaked density profiles have been achieved with a low speed vertical pellet injector located at about mid-radius on the high field side. The performance is comparable to that achieved with a high-speed horizontal pellet injector. Possible reasons for this behaviour are discussed, among them the presence of an 'MHD' drift once particles reach the q = 1 surface. The effect of peaked density profiles on confinement is discussed. Electron ITBs can be produced at high density in FTU with LHCD only and with the combined use of LHCD and ECRH: T_e0 = 6 keV with n_e0 = 1.4 × 10²⁰ m⁻³ and H₉₇ = 1.6. Turbulence is strongly reduced. Ions are heated by collisions with ΔT_i/T_i up to 35% showing that e-ITBs are not degraded by the electron–ion collisions. Particle pinch studies have been made at high densities in full current drive conditions where the Ware pinch plays no role. An anomalous inward pinch exists even at these high densities (n_e0 = 1.5 × 10²⁰ m⁻³). Despite the absence of energetic particles in FTU, MHD spectroscopy has revealed high frequency modes (30–80 kHz) that might have consequences for burning plasmas.

Two experiment campaigns were conducted on the HL-2A tokamak in 2003 and 2004 after the first plasma was obtained at the end of 2002. Progresses in many aspects have been made, especially in the divertor discharge and feedback control of plasma configuration. Up to now, the following operation parameters have been achieved: I_p = 320 kA, B_t = 2.2 T and discharge duration T_d = 1580 ms. With the feedback control of plasma current and horizontal position, an excellent repeatability of the discharge has been achieved. The tokamak has been operated at both limiter configuration and single null (SN) divertor configuration. The HL-2A SN divertor configuration is simulated with the MHD equilibrium code SWEQU. When the divertor configuration is formed, the impurity radiation in the main plasma decreases remarkably.

Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented.

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 10²⁰ m⁻³, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

This paper presents an overview of experimental results and progress made in investigating the role of magnetic configuration on stability and transport in the TJ-II stellarator. Global confinement studies have revealed a positive dependence of energy confinement on the rotational transform and plasma density, together with different parametric dependences for metallic and boronized wall conditions. Spontaneous and biasing-induced improved confinement transitions, with some characteristics that resemble those of previously reported H-mode regimes in other stellarator devices, have been observed. Also, magnetic configuration scan experiments have shown an interplay between magnetic structure (rationals, magnetic shear), transport and electric fields. Although the dc radial electric fields are comparable with those expected from neoclassical calculations, additional mechanisms based on neoclassical/turbulent bifurcations and kinetic effects are needed to explain the impact of magnetic topology on flows and radial electric fields. Hydrocarbon fuelling experiments in configurations with a low-order rational value in the rotational transform located in the proximity of the last closed flux surface (n = 4/m = 2) have shown the impurity screening properties related to the expected divertor effect. The first TJ-II experiments in neutral beam injection plasmas are reported.

Confinement in the reversed field pinch (RFP) has been shown to increase strongly with current profile control. The MST RFP can operate in two regimes: the standard regime with a naturally occurring current density profile, robust reconnection and dynamo activity; and the improved confinement regime with strong reduction in reconnection, dynamo and transport. New results in standard plasmas include the observation of a strong two-fluid Hall effect in reconnection and dynamo, the determination that the m = 0 edge resonant mode is nonlinearly driven, and the determination that tearing modes can lock to the wall via eddy currents in the shell. New results in improved confinement plasmas include observations that such plasmas are essentially dynamo-free, contain several isolated magnetic islands (as opposed to a stochastic field) and contain reduced high frequency turbulence. Auxiliary current drive and heating is now critical to RFP research. In MST, a programme to apply auxiliary systems to the RFP is underway and progress has accrued in several techniques, including lower hybrid and electron Bernstein wave injection, ac helicity injection current drive, pellet injection and neutral beam injection.

Significant theoretical and experimental progress continues to be made at the University of Rochester's Laboratory for Laser Energetics (LLE), charting the path to direct-drive inertial confinement fusion (ICF) ignition. Direct drive offers the potential for higher-gain implosions than x-ray drive and is a leading candidate for an inertial fusion energy power plant. LLE's direct-drive ICF ignition target designs for the National Ignition Facility (NIF) are based on hot-spot ignition. A cryogenic target with a spherical DT-ice layer, within or without a foam matrix, enclosed by a thin plastic shell, will be directly irradiated with ∼1.5 MJ of laser energy. Cryogenic and plastic/foam (surrogate-cryogenic) targets that are hydrodynamically scaled from these ignition target designs are imploded on the 60-beam, 30 kJ, UV OMEGA laser system to validate the key target physics issues, including energy coupling, hydrodynamic instabilities and implosion symmetry. Prospects for direct-drive ignition on the NIF are extremely favourable, even while it is in its x-ray-drive irradiation configuration, with the development of the polar-direct-drive concept. A high-energy petawatt capability is being constructed at LLE next to the existing 60-beam OMEGA compression facility. This OMEGA EP (extended performance) laser will add two short-pulse, 2.6 kJ beams to the OMEGA laser system to backlight direct-drive ICF implosions and study fast-ignition physics with focused intensities up to 6 × 10²⁰ W cm⁻².

The concept of a power plant for fast-ignition heavy ion fusion is developed. It is based on repetitive detonation of a cylindrical direct-drive target, producing 750 MJ of fusion yield in each microexplosion. A heavy-ion driver system providing consequent compression and ignition of the cylindrical DT target is described. Data on energy fluxes generated by the microexplosion are given. The design of the thin liquid wall reactor chamber is presented. The behaviour of the liquid film at the first wall and the blanket coolant and material under a pulsed energy flux loading is analysed. The energy conversion thermal scheme and power plant output parameters are presented. The state of the art at the ITEP-TWAC experimental accelerator is described.