Table of contents

Recent JT-60U results leading to high integrated performance are reported with emphasis on the projection to the reactor-relevant regime. Negative-ion-based neutral beam (NB) and electron cyclotron (EC) power increased up to 6.2 MW and 3 MW, respectively. A high β_p H-mode plasma with full non-inductive current drive has been obtained at 1.8 MA, and the fusion triple product reached 3.1 × 10²⁰ m⁻³ keV s. A high beta with β_N = 2.7 was maintained for 7.4 s. Neoclassical tearing mode suppression with EC was accomplished using a real-time feedback control system, and improvement in β_N was obtained. The stable existence of a current hole was observed. A high DT-equivalent fusion gain of 0.8 was maintained for 0.55 s in a plasma with a current hole. Current profile control in high bootstrap current reversed shear plasmas was demonstrated using N-NB and LH. A new operation scenario has been established in which a plasma with a high bootstrap current fraction and internal transport barriers (ITBs) is produced without the use of an OH coil. An ECCD study was undertaken in a reactor-relevant high T_e regime. A new type of Alfvén eigenmode mode has been proposed and has been found to explain the observed frequency chirp quite well. High confinement reversed shear plasmas with T_e > T_i were obtained. Ar exhaust with EC heating was obtained in a high β_p mode plasma. Impurity accumulation related to strong ITBs in a reversed shear plasma and degradation of ITB by ECH in a weak positive shear plasma have been found. Dedicated measurement of edge localized mode dynamics and scrape-off-layer plasma flow have advanced understanding of the physics. N-NB heating in an Ar-seed plasma extended the density region to 95% of the Greenwald density, with HH_y2 = 0.9. Enhancement of pedestal pressure was obtained with an increase in β_p in a high triangularity configuration.

Scientific and technical activities on JET focus on the issues likely to affect the ITER design and operation. Our understanding of the ITER reference mode of operation, the ELMy H-mode, has progressed significantly. The extrapolation of ELM size to ITER has been re-evaluated. Neoclassical tearing modes have been shown to be meta-stable in JET, and their beta limits can be raised by destabilization (modification) of sawteeth by ion cyclotron radio frequency heating (ICRH). Alpha simulation experiments with ICRH accelerated injected ⁴He beam ions provide a new tool for fast particle and magnetohydrodynamic studies, with up to 80–90% of plasma heating by fast ⁴He ions. With or without impurity seeding, a quasi-steady-state high confinement (H₉₈ = 1), high density (n_e/n_GW = 0.9–1) and high β (β_N = 2) ELMy H-mode has been achieved by operating near the ITER triangularity (δ∼ 0.40–0.5) and safety factor (q₉₅∼ 3), at Z_eff∼ 1.5–2. In advanced tokamak (AT) scenarios, internal transport barriers (ITBs) are now characterized in real time with a new criterion, . Tailoring of the current profile with lower hybrid current drive provides reliable access to a variety of q profiles, lowering access power for barrier formation. Rational q surfaces appear to be associated with ITB formation. Alfvén cascades were observed in reversed shear plasmas, providing identification of q profile evolution. Plasmas with 'current holes' were observed and modelled. Transient high confinement AT regimes with H₈₉ = 3.3, β_N = 2.4 and ITER-relevant q < 5 were achieved with reversed magnetic shear. Quasi-stationary ITBs are developed with full non-inductive current drive, including ∼50% bootstrap current. A record duration of ITBs was achieved, up to 11 s, approaching the resistive time. For the first time, pressure and current profiles of AT regimes are controlled by a real-time feedback system, in separate experiments. Erosion and co-deposition studies with a quartz micro-balance show reduced co-deposition. Measured divertor thermal loads during disruptions in JET could modify ITER assumptions.

The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last international atomic energy agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: (1) we have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, we have achieved β_NH₈₉ ⩾ 10 for 4τ_E limited by the neoclassical tearing mode (NTM); (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m, n) = (3, 2) NTM and then increased β_T by 60%; (4) we have produced ECCD stabilization of the (2, 1) NTM in initial experiments; (5) we have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) we have demonstrated stationary tokamak operation for 6.5 s (36τ_E) at the same fusion gain parameter of as ITER but at much higher q₉₅ = 4.2. We have developed general improvements applicable to conventional and AT operating modes: (1) we have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, edge localized modes (ELM) heat load to the divertor and which can run for long periods of time (3.8 s or 25τ_E) with constant density and constant radiated power; (2) we have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) we have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. We have made detailed investigations of the edge pedestal and scrape-off layer (SOL): (1) atomic physics and plasma physics both play significant roles in setting the width of the edge density barrier in H-mode; (2) ELM heat flux conducted to the divertor decreases as density increases; (3) intermittent, bursty transport contributes to cross field particle transport in the SOL of H-mode and, especially, L-mode plasmas.

Recent results from the ASDEX Upgrade experimental campaigns 2001 and 2002 are presented. An improved understanding of energy and particle transport emerges in terms of a 'critical gradient' model for the temperature gradients. Coupling this to particle diffusion explains most of the observed behaviour of the density profiles, in particular, the finding that strong central heating reduces the tendency for density profile peaking. Internal transport barriers (ITBs) with electron and ion temperatures in excess of 20 keV (but not simultaneously) have been achieved. By shaping the plasma, a regime with small type II edge localized modes (ELMs) has been established. Here, the maximum power deposited on the target plates was greatly reduced at constant average power. Also, an increase of the ELM frequency by injection of shallow pellets was demonstrated. ELM free operation is possible in the quiescent H-mode regime previously found in DIII-D which has also been established on ASDEX Upgrade. Regarding stability, a regime with benign neoclassical tearing modes (NTMs) was found. During electron cyclotron current drive (ECCD) stabilization of NTMs, β_N could be increased well above the usual onset level without a reappearance of the NTM. Electron cyclotron resonance heating and ECCD have also been used to control the sawtooth repetition frequency at a moderate fraction of the total heating power. The inner wall of the ASDEX Upgrade vessel has increasingly been covered with tungsten without causing detrimental effects on the plasma performance. Regarding scenario integration, a scenario with a large fraction of noninductively driven current (⩾50%), but without ITB has been established. It combines improved confinement (τ_E/τ_ITER98 ≈ 1.2) and stability (β_N ⩽ 3.5) at high Greenwald fraction (n_e/n_GW ≈ 0.85) in steady state and with type II ELMy edge and would offer the possibility for long pulses with high fusion power at reduced current in ITER.

Operation of Tore Supra has resumed after a major upgrade of internal components in order to increase the heat extraction capability to 25 MW for 1000 s and address long-pulse research at the level of power density flowing through the last closed flux surface relevant for the next step. This overview highlights first the related technology developments: industrial realization and tests with plasma of about 600 limiter components made of some 12 000 actively cooled CFC tiles. World record breaking discharges with a coupled (and extracted) energy of 0.74 GJ have been obtained, with durations exceeding 4 min, limited only by the long-pulse capability of the heating and current drive systems. Particle balance analysis shows that the in-vessel deuterium inventory never reached saturation. For the same plasma parameters, the in-vessel inventory is reduced by a factor of ≈2 when efficient supersonic gas injection is used. New physics is observed when the loop voltage vanishes for long times: the density remains peaked in spite of the absence of the Ware pinch effect, and, independently, the response to additional heating becomes dominated by a complex interplay between transport and current drive. The contributions of the EURATOM-CEA Association to fusion energy research also include new developments in the field of plasma-facing components, negative ion beam injection (1000 s H⁻ source and SINGAP acceleration to 1 MeV) and superconductors (cable in conduit concept, connections, testing). Finally, European studies for Cadarache as a site for ITER conclude that all ITER technical requirements are met.

An overview of steady state tokamak studies in TRIAM-1M (R₀ = 0.8 m, a × b = 0.12 m × 0.18 m and B = 8 T) is presented. The current ramp-up scenario without using centre solenoid coils is reinvestigated with respect to controllability of the current ramp-up rate at the medium density region of (1–2) × 10¹⁹ m⁻³. The plasma is initiated by ECH (fundamental o-mode at 170 GHz with 200 kW) at B = 6.7 T, and the ramp-up rate below the technical limit of 150 kA s⁻¹ for ITER can be achieved by keeping the LH power less than 100 kW during the current ramp-up phase. The physics understanding of the enhanced current drive (ECD) mode around the threshold power level has progressed from a viewpoint of transition probability. A transition frequency, f_trans, for the ECD transition is determined as a function of P_CD. At ∼70 kW no transition occurs for an f_trans value of ∼0.017 Hz, meaning almost zero transition probability. With increasing P_CD > P_th, f_trans increases up to 10 Hz, and the transition tends to occur with high probability. The record value of the discharge duration is updated to 3 h 10 min in a low and low power (<10 kW) discharge. The global particle balance in long duration discharges is investigated, and the temporal change in wall pumping rate is determined. Although the density was low, the gas supply had to be stopped at 30 min after the plasma initiation to maintain the density constant. After that the density was sustained by the recycling flux alone until the end of the discharges. In addition to the recycling problem, in the high power and high density experiments, the localized PWI affects the SSO of the tokamak plasma. The effects of enhanced influx of metal impurities (Fe, Cr, Ni, Mo) on sustainment of the high performance ECD plasma are investigated. In order to evaluate the helium bombarding effects on the plasma facing component and hydrogen recycling in the future burning plasma, microscopic damage of metals exposed to long duration helium discharges was studied. The total exposure time was 128 s. From thermal desorption experiments for the specimens the amount of retained helium was evaluated as 3.9 × 10²⁰ He m⁻² and the scale length to be ∼1 mm in the SOL.

Research on the Alcator C-Mod tokamak [1] is focused on high particle- and power-density plasma regimes to understand particle and energy transport in the core, the dynamics of the H-mode pedestal, and scrape-off layer and divertor physics. The auxiliary heating is provided exclusively by RF waves, and both the physics and technology of RF heating and current drive are studied. The momentum which is manifested in strong toroidal rotation, in the absence of direct momentum input, has been shown to be transported in from the edge of the plasma following the L–H transition, with timescale comparable to that for energy transport. In discharges which develop internal transport barriers, the rotation slows first inside the barrier region, and then subsequently outside of the barrier foot. Heat pulse propagation studies using sawteeth indicate a very narrow region of strongly reduced energy transport, located near r/a = 0.5. Addition of on-axis ICRF heating arrests the buildup of density and impurities, leading to quasi-steady conditions. The quasi-coherent mode associated with enhanced D-Alpha (EDA) H-mode appears to be due to a resistive ballooning instability. As the pedestal pressure gradient and temperature are increased in EDA H-mode, small ELMs appear; detailed modelling indicates that these are due to intermediate n peeling–ballooning modes. Phase contrast imaging has been used to directly detect density fluctuations driven by ICRF waves in the core of the plasma, and mode conversion to an intermediate wavelength ion cyclotron wave has been observed for the first time. The bursty turbulent density fluctuations, observed to drive rapid cross-field particle transport in the edge plasma, appear to play a key role in the dynamics of the density limit. Preparations for quasi-steady-state advanced tokamak studies with lower hybrid current drive are well underway, and time dependent modelling indicates that regimes with high bootstrap fraction can be produced.

The Tokamak à Configuration Variable (TCV) tokamak (R = 0.88 m, a < 0.25 m, B < 1.54 T) programme is based on flexible plasma shaping and heating for studies of confinement, transport, control and power exhaust. Recent advances in fully sustained off-axis electron cyclotron current drive (ECCD) scenarios have allowed the creation of plasmas with high bootstrap fraction, steady-state reversed central shear and an electron internal transport barrier. High elongation plasmas, κ = 2.5, are produced at low normalized current using far off-axis electron cyclotron heating and ECCD to broaden the current profile. Third harmonic heating is used to heat the plasma centre where the second harmonic is in cut-off. Both second and third harmonic heating are used to heat H-mode plasmas, at the edge and centre, respectively. The ELM frequency is decreased by the additional power. In separate experiments, the ELM frequency can be affected by locking to an external perturbation current in the internal coils of TCV. Spatially resolved current profiles are measured at the inner and outer divertor targets by Langmuir probe arrays during ELMs. The strong, reasonably balanced currents are thought to be thermoelectric in origin.

An overview of the Frascati tokamak upgrade (FTU) results during the period 2000–2002 is presented. Long duration internal transport barriers (ITBs) have been obtained on FTU with combined injection of lower hybrid (LH) and electron cyclotron (EC) waves in 5 T/0.5 MA discharges. The ITB phase lasts about 10 energy confinement times and is characterized by an energy confinement time up to 1.6 times the ITER97 L-mode scaling. Temperatures up to 11 keV are measured at 0.9 × 10²⁰ m⁻³ central density. ITB studies using IBW injection have also been continued up to 8 T/0.8 MA. The LH system has operated at full power allowing us to complete the current drive (CD) studies at ITER-relevant densities. At these density values the electrons and ions are coupled and an increase in the ion temperature is clearly observed. A preliminary sign of enhanced CD efficiency has been obtained in the combined injection of EC and LH waves at magnetic field values lower than the resonant field for thermal EC absorption. Pellet optimization studies have been performed in order to test the conditions under which a quasi-steady-state confinement improvement can be obtained and impurity accumulation can be avoided. The confinement time in Ohmic discharges is generally in agreement with the ITER97 L-mode scaling. Transient confinement improvement is observed for a duration less than one energy confinement time. Radiation improved mode studies have been started thanks to the recently inserted boronization system, which has allowed reduction of the radiated power. Confinement improvement with neon injection has been observed in 6 T/0.9 MA discharges. Transport studies on profile stiffness and magnetohydrodynamic (MHD) studies of fast reconnection and snakes will also be presented.

Transitions to the improved confinement regimes were studied in the T-10 tokamak with electron cyclotron resonance heating (ECRH). The H-mode regime with and without internal transport barrier (ITB) was obtained and studied. The H-mode was obtained by power increase, pellet injection and biasing. The confinement time in the pellet enhanced confinement regime is 30% higher than that in the spontaneous H-mode. In biasing experiments an enhancement factor of up to 1.55 was obtained for the energy confinement time. It was shown that subtle changes in the safety factor q profile allow us to obtain either ITB or external transport barrier only, or both of them, other conditions being approximately similar. The global confinement in the regime with the density above the Greenwald limit under strong gas-puffing and ECRH was investigated. To identify the type of turbulence responsible for transport, the Ohmic and ECRH regimes with variation of safety factor at the limiter q_L and gas puffing were considered. Two regions were observed along the minor radius with different turbulence properties.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m⁻² has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

The low aspect ratio of the mega amp spherical tokamak (MAST) allows differentiation between different forms of the H-mode threshold scaling. With optimized fuelling using inboard puffing, and a connected double null divertor (DND) magnetic configuration, the H-mode power threshold data lie about 1.7 times higher than recent scaling laws. Slight magnetic configuration changes, of the order of the ion Larmor radius, around a connected DND significantly influence H-mode access. H-mode confinement in discharges with low frequency edge localized modes (ELMs) is generally consistent with international scaling laws, e.g. IPB98(y,2). Strong indications of both particle and energy internal transport barriers have been seen. Normalized beta values β_N > 5 have been obtained, approaching the ideal n = 1 no wall external kink stability limit. Sawtooth triggered neo-classical tearing modes have been observed; numerical modelling of the island evolution reproduces mode behaviour well and confirms the significance of stabilizing field curvature effects. Divertor power loading studies, including transient effects due to ELMs, show a strong bias of power efflux to the outboard targets, where it is more easily handled. ELM energy losses, Δ W_ELM, are less than 4% of the stored energy in all regimes explored so far, but ELM effluxes extending 30 cm outside the outboard separatrix have been measured. Toroidally asymmetric divertor biasing resulted in significant broadening of the D_α profile on the biased components and a reduction in the total power to the unbiased components. Halo current magnitudes and asymmetries are generally small compared with conventional tokamaks; recent measurements show that the plasma behaves more as a voltage source than a current source. Initial neutral beam current drive experiments indicate non-inductively driven current values (I_NBI ∼ 0.3I_p) comparable with code predictions.

In the first four years of the LHD experiment, several encouraging results have emerged, the most significant of which is that MHD stability and good transport are compatible in the inward shifted axis configuration. The observed energy confinement at this optimal configuration is consistent with ISS95 scaling with an enhancement factor of 1.5. The confinement enhancement over the smaller heliotron devices is attributed to the high edge temperature. We find that the plasma with an average beta of 3% is stable in this configuration, even though the theoretical stability conditions of Mercier modes and pressure driven low-n modes are violated. In the low density discharges heated by NBI and ECR, internal transport barrier (ITB) and an associated high central temperature (>10 keV) are seen. The radial electric field measured in these discharges is positive (electron root) and expected to play a key role in the formation of the ITB. The positive electric field is also found to suppress the ion thermal diffusivity as predicted by neoclassical transport theory. The width of the externally imposed island is found to decrease when the plasma is collisionless with finite beta and increase when the plasma is collisional. The ICRF heating in LHD is successful and a high energy tail (up to 500 keV) has been detected for minority ion heating, demonstrating good confinement of the high energy particles. The magnetic field line structure unique to the heliotron edge configuration is confirmed by measuring the plasma density and temperature profiles on the divertor plate. A long pulse (2 min) discharge with an ICRF power of 0.4 MW has been demonstrated and the energy confinement characteristics are almost the same as those in short pulse discharges.

Energy confinement comparable with tokamak quality is achieved in the Madison Symmetric Torus (MST) reversed field pinch (RFP) at a high beta and low toroidal magnetic field. Magnetic fluctuations normally present in the RFP are reduced via parallel current drive in the outer region of the plasma. In response, the electron temperature nearly triples and beta doubles. The confinement time increases ten-fold (to ∼10 ms), which is comparable with L- and H-mode scaling values for a tokamak with the same plasma current, density, heating power, size and shape. Runaway electron confinement is evidenced by a 100-fold increase in hard x-ray bremsstrahlung. Fokker–Planck modelling of the x-ray energy spectrum reveals that the high energy electron diffusion is independent of the parallel velocity, uncharacteristic of magnetic transport and more like that for electrostatic turbulence. The high core electron temperature correlates strongly with a broadband reduction of resonant modes at mid-radius where the stochasticity is normally most intense. To extend profile control and add auxiliary heating, rf current drive and neutral beam heating are in development. Low power lower-hybrid and electron Bernstein wave injection experiments are underway. Dc current sustainment via ac helicity injection (sinusoidal inductive loop voltages) is also being tested. Low power neutral beam injection shows that fast ions are well-confined, even in the presence of relatively large magnetic fluctuations.

A coordinated, focused effort is underway to develop Laser Inertial Fusion Energy. The key components are developed in concert with one another and the science and engineering issues are addressed concurrently. Recent advances include: target designs have been evaluated that show it could be possible to achieve the high gains (>100) needed for a practical fusion system.These designs feature a low-density CH foam that is wicked with solid DT and over-coated with a thin high-Z layer. These results have been verified with three independent one-dimensional codes, and are now being evaluated with two- and three-dimensional codes. Two types of lasers are under development: Krypton Fluoride (KrF) gas lasers and Diode Pumped Solid State Lasers (DPSSL). Both have recently achieved repetitive 'first light', and both have made progress in meeting the fusion energy requirements for durability, efficiency, and cost. This paper also presents the advances in development of chamber operating windows (target survival plus no wall erosion), final optics (aluminium at grazing incidence has high reflectivity and exceeds the required laser damage threshold), target fabrication (demonstration of smooth DT ice layers grown over foams, batch production of foam shells, and appropriate high-Z overcoats), and target injection (new facility for target injection and tracking studies).

This article summarizes the papers that are related to the theory of magnetic confinement plasmas.

This paper summarizes the results presented at the IAEA Fusion Energy Conference 2002 with respect to the performance and confinement of tokamaks, spherical tokamaks, stellarators, reversed field pinches and mirrors. Transport, internal transport barriers, methods to control pressure and current density profiles by auxiliary heating and non-inductive current drive will be discussed and compared to the predictions for ITER. Based on the results presented at this conference it can be concluded that one has confidence that the ITER performance specifications will be achieved in the future. Marked progress on alternative confinement concepts can be noted as well.

Based on the now well-known and experimentally observed critical gradient length (L_Te) in tokamaks, we present a continuum one-dimensional model for explaining self organized heat transport in tokamaks. The key parameters of this model include a hysteresis parameter, which ensures that the switching of the heat transport coefficient χ upwards and downwards takes place at two different values of R/L_Te. Extensive numerical simulations of this model reproduce many features of present day tokamaks, such as sub-marginal temperature profiles, intermittent transport events, 1/f scaling of the frequency spectra, propagating fronts, etc. This model utilizes a minimal set of phenomenological parameters, which may be determined from experiments and/or simulations. Analytical and physical understanding of the observed features has also been attempted.

The steady state superconducting tokamak (SST-1) is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (κ) and triangularity (δ). Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. A NbTi based cable-in-conduit conductor (CICC) has been fabricated by M/S Hitachi Cables Ltd., Japan under specification and supervision of the Institute for Plasma Research (IPR). The suitability of this CICC for the SST-1 magnets has been validated through test carried out on a model coil wound from this CICC. Toroidal and poloidal SC magnets have been fabricated and factory acceptance tests have been performed. SC magnets require liquid helium (LHe) cooled current leads, electrical isolators at LHe temperature, superconducting bus bars and LHe transfer lines. Full scale prototypes of these have been developed and tested successfully. SC magnets will be cooled to 4.5 K by forced flow of supercritical helium through the CICC. A 1 kW grade liquefier/refrigerator has been installed and is in final stages of commissioning at IPR. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors (VSs) having ports and 16 rings with -shaped cross-section. To establish the fabrication methodology for this, a fullscale prototype of the vessel with two VSs and three rings has been fabricated and tested successfully. Based on this the fabrication of the VSs and rings is in final stage of fabrication. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. SST-1 will have three different high power radio frequency systems to additionally heat and non-inductively drive plasma current to sustain the plasma in steady state for a duration of up to 1000 s. Ion cyclotron resonance frequency (ICRF) and electron cyclotron resonance frequency (ECRF) systems will primarily be used for heating the plasma while lower hybrid waves will be used for non inductive lower hybrid current drive (LHCD). A neutral beam injection with peak power of 0.8 MW with variable beam energy in range of 10–80 keV will be used as additional auxiliary heating system. A number of prototypes for various critical components have confirmed the fabrication methodology. The fabrication of most of the subsystems is nearing completion and many components have already been accepted on site. Erection and installation of the base of the mechanical structure has already been initiated in the SST hall. This paper reports on the results of the tests on various prototypes and actual components to be used on SST-1 for various subsystems.

Coupled theoretical and computational work is presented aimed at understanding and modelling stimulated Raman backscattering (SRBS) relevant to laser–plasma interactions in large-scale, homogeneous plasmas. With the aid of a new code for simulating and studying the nonlinear coupling in space–time of a large number of modes, and an Eulerian Vlasov–Maxwell code for studying the evolution of large amplitude electron plasma waves, we report results and their interpretations to elucidate the following five observed, nonlinear phenomena associated with SRBS: coupling of SRBS to Langmuir decay instabilities (LDIs); effect of ion-acoustic damping on SRBS; cascading of LDI; stimulated Raman scattering cascades; and stimulated electron acoustic wave scattering (SEAS).

The prediction of major disruptions caused by the density limit, the plasma current ramp-down with high internal inductance l_i, the low density locked mode and the β-limit has been investigated in JT-60U. The concept of 'stability level', newly proposed in this paper to predict the occurrence of a major disruption, is calculated from nine input parameters every 2 ms by the neural network and the start of a major disruption is predicted when the stability level decreases to a certain level, the 'alarm level'. The neural network is trained in two steps. It is first trained with 12 disruptive and six non-disruptive shots (total of 8011 data points). Second, the target output data for 12 disruptive shots are modified and the network is trained again with additional data points generated by the operator. The 'neural-net disruption predictor' obtained has been tested for 300 disruptive shots (128 945 data points) and 1008 non-disruptive shots (982 800 data points) selected from nine years of operation (1991–1999) of JT-60U. Major disruptions except for those caused by the β-limit have been predicted with a prediction success rate of 97–98% at 10 ms prior to the disruption and higher than 90% at 30 ms prior to the disruption while the false alarm rate is 2.1% for non-disruptive shots. This prediction performance has been confirmed for 120 disruptive shots (56 163 data points), caused by the density limit, as well as 1032 non-disruptive shots (1004 611 data points) in the last four years of operation (1999–2002) of JT-60U. A careful selection of the input parameters supplied to the network and the newly developed two-step training of the network have reduced the false alarm rate resulting in a considerable improvement of the prediction success rate.

A confinement database of the reversed-field pinch (RFP) devices of the toroidal pinch experiment (TPE) series has been established. The TPE database contains global confinement properties under various experimental conditions for four different devices and seven different cases, which cover two decades of experiments. Characteristics of the TPE database have been extracted under consistent experimental conditions of conventional RFP discharges, and scaling laws of the poloidal beta, β_p and energy confinement time, τ _E, are given against independent parameters that include the size of the plasma and the pinch parameter Θ. Scaling laws of and are obtained for a set of scans, where a is the plasma minor radius, I_p the plasma current and N the column density. The scaling of τ _E on a, I_p, and I/N is similar to that dominated by the transport under tearing instability. Another scaling shows that τ _E is inversely proportional to the Joule input power. The scaling for the apparent effective charge shows a positive correlation both with I_p and I/N.

A new combined method for investigation of magnetohydrodynamic (MHD) activity in fusion experiments is proposed. The main advantages of this approach are the possibility of a direct comparison of theory predictions with experimental observations by the simultaneous use of several diagnostics. The method has been applied successfully to different types of instabilities in conventional and advanced tokamak scenarios in ASDEX Upgrade. As examples, double tearing mode as well as neoclassical tearing modes are considered. The combined use of different diagnostics allows an accurate determination of the mode's eigenfunction.

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks, leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments (Scoville J.T. and La Haye R.J. 2003 Nucl. Fusion43 250) and resistive wall mode experiments (Garofalo A.M., La Haye R.J. and Scoville J.T. 2002 Nucl. Fusion42 1335) done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made in DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported (La Haye R.J. and Scoville J.T. 1991 Rev. Sci. Instrum.61 2146). Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the Ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all the coils are one-third the values indicated from the stability experiments (Garofalo et al 2002, Scoville and La Haye 2003). These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.

Over the last two years, several experiments relevant for the study of particle transport and density profile evolution, have been performed at JET. They can be classified as stationary discharges with and without central particle source due to the beams, quasi-stationary discharges with deuterium gas puffing, deep pellet fuelled discharges and discharges perturbed by cold pulses obtained by shallow pellet injection. All these experimental scenarios have been simulated by means of the JETTO transport code, employing different transport models: purely empirical models and the semi-empirical mixed Bohm/gyro-Bohm transport model, both with the addition of different theory-based expressions for the anomalous particle pinch and the first principle Weiland transport model. The coefficients used to scale the pinch velocity in the purely empirical and in the mixed Bohm/gyro-Bohm model have been varied from shot to shot. In this paper, the results of the simulations are presented. The main conclusions are that, for the cases studied in this paper, the sawtooth activity is the main particle transport mechanism in the plasma centre (r/a ⩽ 0.5). Nevertheless, to reproduce the density profile in the gradient zone (0.5 ⩽ r/a ⩽ 0.9), an anomalous pinch seems to be necessary, at least for L-mode plasmas. This anomalous convective flux is well reproduced by the off-diagonal elements of the transport matrix given by the Weiland model.

This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed.

By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study.

In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was successfully fabricated. It withstood the high heat flux test at 2.7 MW m⁻². Also, a correlation parameter of the Li₂TiO₃ pebble bed made by the sol–gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data.

The GLF23 and multi-mode core transport models are used along with models for the H-mode pedestal to predict the fusion performance for the International Thermonuclear Experimental Reactor, Fusion Ignition Research Experiment, and IGNITOR tokamak designs. Simulations using combinations of core and pedestal models have also been compared with experimental data for H-mode profiles in DIII-D, JET, and Alcator C-Mod. Power-independent (ballooning mode limit) and power-dependent pedestal scalings lead to very different predictions when used with the core models. Although the two drift-wave transport models reproduce the core profiles in a wide variety of tokamak discharges, they differ in their projections to burning plasma experiments for the same pedestal parameters. Differences in the core transport models in their response to the ion temperature gradient (i.e. their stiffness) and impact of the power dependence of the H-mode pedestal on fusion performance predictions are discussed.

We report the results of an experimental and theoretical international project dedicated to the study of quasi-single helicity (QSH) reversed field pinch (RFP) plasmas. The project has involved several RFP devices and numerical codes. It appears that QSH spectra are a robust feature common to all the experiments. Our results expand and reinforce the evidence that the formation of self-organized states with one dominant helical mode (Ohmic SH state) is an approach complementary to that of active control of magnetic turbulence to improve confinement in a steady state RFP.

The PDF file provided contains web links to all articles in this volume.

The first 17 articles published in this issue of Nuclear Fusion form a special section with the aim of monitoring the progress made in experiments on fusion physics that have been conducted worldwide up to the end of 2002. These articles are based on overview reports from the various experimental teams presented at the Fusion Energy Conference (FEC-2002). This conference was organized by the IAEA together with the French host organization CEA and was held in Lyon, France, in October 2002. During 2003 the conference overviews have been rewritten and extended for the purpose of this special section and submitted to our standard double-referee peer review. While not all teams have made use of this opportunity, a sufficiently large number have, so that this issue presents a reasonably complete picture of the progress made since the FEC-2000 was held in Sorrento.

The articles are arranged in the following sequence:

Tokamaks:

Performance: JT-60U, JET, DIII-D, ASDEX-U;

Steady state/long pulse operation: Tore Supra, TRIAM;

Heating methods: C-Mod (ICRH), TCV and T-10 (ECRH) and FTU (LHH + ECRH).

Alternative magnetic confinement concepts:

Stellarators: LHD;

Reversed field pinches: MST.

Inertial confinement:

Direct drive.

Two of the five summaries presented at the Lyon conference, which aim to report on the progress made in fusion physics, complete this special section. These summaries are structured along the lines of specific physics issues instead of specific devices: `Theory of magnetic confinement' and `Experiments in confinement and wave--plasma interaction'.

I hope that this special section will give the interested reader an up-to-date picture of the progress made in nuclear fusion research.