Table of contents

The focus of effort in ITER EDA since 1998 has been on the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. Drawing on the design solutions already developed, and using the latest physics results and outputs from technology R&D projects, the Joint Central Team and Home Teams, working together, have been able to progress towards a new design which will allow the exploration of a range of burning plasma conditions, with a capacity to progress towards possible modes of steady state operation. The new ITER design, whilst having reduced technical objectives from those of its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. The main features of the current design and of its projected performance are introduced and the outlook for construction and operation is summarized.

With the main aim of providing the physics basis for ITER and steady state tokamak reactors, JT-60U has been optimizing operational concepts and extending discharge regimes towards simultaneous sustainment of high confinement, high β_N, high bootstrap fraction, full non-inductive current drive, and efficient heat and particle exhaust utilizing a variety of heating, current drive, torque input and particle control capabilities. In the two advanced operation regimes (reversed magnetic shear (RS mode) and weak magnetic shear (high β_p mode)), ELMy H mode discharges characterized by both internal and edge transport barriers and by high bootstrap current fraction f_BS have been sustained near the steady state current profile solutions under full non-inductive current drive with appropriate driven current profiles (high β_p: HH_y2 ≈ 1.4 and β_N ≈ 2.5 with negative ion based neutral beams (N-NBs); RS: HH_y2 ≈ 2.2 and β_N ≈ 2 with f_BS ≈ 80%). Multiple pellet injection has extended the density region with high confinement. These operational modes have been extended to the reactor relevant regime with small values of collisionality and normalized gyroradius, and T_e∼T_i. In the RS regime, Q_DT^eq = 0.5 has been sustained for 0.8 s. Stability has been improved in the high β_p mode regime by suppression of the neoclassical tearing mode with local ECCD and in the RS mode regime by wall stabilization. The internal transport barrier structure has been controlled by toroidal rotation profile modification, and transport studies have revealed a semiglobal dynamics of the internal transport barrier structure. Simultaneous pumping at the inner and the outer divertor leg has enhanced He exhaust by ∼40%. Argon puff experiments have improved confinement at high density with detached divertor owing to high pedestal temperature T_i^ped. In H modes, the core confinement degraded with decreasing T_i^ped, suggesting stiff core profiles. An operational region of grassy ELMs with small divertor heat load has been established at high triangularity, high q₉₅ and high β_p. A record value of the neutral beam current drive efficiency of 1.55 × 10¹⁹A m^-2 W^-1 has been demonstrated using N-NB. Abrupt large amplitude events causing a drop in neutron emission have been discovered with frequency inside the TAE gap. Disruption studies have clarified that runaway current is terminated by MHD fluctuations when the surface q decreases to 2-3.

The JET experimental campaign has focused on studies in support of the ITER physics basis. An overview of the results obtained is given for the reference ELMy H mode and advanced scenarios, which in JET are based on internal transport barriers. JET studies for ELMy H mode have been instrumental in the definition of ITER FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, but the observed density scaling fits a two term (core and edge) model better. Significant progress in neoclassical tearing mode limits has been made showing that ITER operation with q₉₅ around 3.3 seems to be optimized. Effective helium pumping and divertor enrichment is found to be well within ITER requirements. Target asymmetries and hydrogen isotope retention are well simulated by modelling codes taking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling effectiveness have been made with the new high field pellet launch facility. Good progress has been made on scenarios for achieving good confinement at high densities, both with radiation improved modes and with high field side pellets. Significant development of advanced scenarios, in view of their application to ITER, has been achieved. Progress towards integrated advanced scenarios is well developed with edge pressure control (impurity radiation). An access domain has been explored showing, in particular, that the power threshold increases with magnetic field but can be significantly reduced when lower hybrid current drive is used to produce target plasmas with negative shear. The role of ion pressure peaking on MHD has been well documented. Lack of sufficient additional heating power and interaction with the septum at high beta prevents assessment of the beta limits (steady plasmas achieved with β_N up to 2.6). Plasmas with a non-inductive current (I_NI/I_p = 60%), well aligned with the plasma current, high beta and good confinement have also been obtained.

The goals of DIII-D advanced tokamak (AT) experiments are investigation and optimization of the upper limits of energy confinement and MHD stability in a tokamak plasma, and simultaneous maximization of the fraction of non-inductive current drive. Significant overall progress has been made in the past two years, as the performance figure of merit β_N H_89P of 9 has been achieved in ELMing H mode for over 16τ_E without sawteeth. The tokamak was also operated at β_NH ≈ 7 for over 35 τ_E or 3τ_R, with the duration limited by the hardware. Real time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike and X points, double/single null balance) and particle control (n_e/n_GW ≈ 0.3, Z_eff < 2.0) were necessary for the long pulse results. A new quiescent double barrier (QDB) regime with simultaneous inner and edge transport barriers and no ELMs has been discovered with a β_N H_89P of 7. The QDB regime has been obtained to date only with counter NBI. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high triangularity plasma shapes. Discharges at n_e/n_GW ≈ 1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real time feedback control of resistive wall modes with external coils and control of neoclassical tearing modes with ECCD.

During the first two years of the LHD experiment the following results have been achieved: (i) higher T_e (T_e(0) = 4.4 keV at ⟨n_e⟩ = 5.3 × 10¹⁸ m^-3 and P_abs = 1.8 MW); (ii) higher confinement (τ_E = 0.3 s, T_e(0) = 1.1 keV at ⟨n_e⟩ = 6.5 × 10¹⁹ m^-3 and P_abs = 2.0 MW); (iii) higher stored energy, W_p^dia = 880 kJ at B = 2.75 T. High performance plasmas have been realized in the inward shifted magnetic axis configuration (R = 3.6 m) where helical symmetry is recovered and the particle orbit properties are improved by a trade-off of MHD stability properties due to the appearance of a magnetic hill. Energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 by up to a factor of 1.6 and was comparable with the ELMy H mode confinement capability in tokamaks. This confinement improvement is attributed to configuration control (inward shift of the magnetic axis) and to the formation of a high edge temperature. The average beta value achieved reached 2.4% at B = 1.3 T, the highest beta value ever obtained in a helical device, and so far no degradation of confinement by MHD phenomena has been observed. The inward shifted configuration has also led to successful ICRF minority ion heating. ICRF powers up to 1.3 MW were reliably injected into the plasma without significant impurity contamination, and a plasma with a stored energy of 200 kJ was sustained for 5 s by ICRF alone. As another important result, long pulse discharges of more than 1 min were successfully achieved separately with an NBI heating of 0.5 MW and with an ICRF heating of 0.85 MW.

Ion and electron temperature profiles in conventional L and H mode on ASDEX Upgrade are generally stiff and limited by a critical temperature gradient length ∇T/T as given by ion temperature gradient (ITG) driven turbulence. ECRH experiments indicate that electron temperature (T_e) profiles are also stiff, as predicted by electron temperature gradient turbulence with streamers. Accordingly, the core and edge temperatures are proportional to each other and the plasma energy is proportional to the pedestal pressure for fixed density profiles. Density profiles are not stiff, and confinement improves with density peaking. Medium triangularity shapes (δ<0.45) show strongly improved confinement up to the Greenwald density n_GW and therefore higher βvalues, owing to increasing pedestal pressure, and H mode density operation extends above n_GW. Density profile peaking at n_GW was achieved with controlled gas puffing rates, and first results from a new high field side pellet launcher allowing higher pellet velocities are promising. At these high densities, small type II ELMs provide good confinement with low divertor power loading. In advanced scenarios the highest performance was achieved in the improved H mode with H_L-89Pβ_N ≈ 7.2 at δ = 0.3 for five confinement times, limited by neoclassical tearing modes (NTMs) at low central magnetic shear (q_min ≈ 1). The T profiles are still governed by ITG and trapped electron mode (TEM) turbulence, and confinement is improved by density peaking connected with low magnetic shear. Ion internal transport barrier (ITB) discharges - mostly with reversed shear (q_min>1) and L mode edge - achieved H_L-89P ⩽ 2.1 and are limited to β_N ⩽ 1.7 by internal and external ideal MHD modes. Turbulence driven transport is suppressed, in agreement with the E × B shear flow paradigm, and core transport coefficients are at the neoclassical ion transport level, where the latter was established by Monte Carlo simulations. Reactor relevant ion and electron ITBs with T_e ≈ T_i ≈ 10 keV were achieved by combining ion and electron heating with NBI and ECRH, respectively. In low current discharges full non-inductive current drive was achieved in an integrated high performance H mode scenario with [`n]_e = n_GW, high β_p = 3.1, β_N = 2.8 and H_L-89P = 1.8, which developed ITBs with q_min ≈ 1. Central co-ECCD at low densities allows a high current drive fraction of >80%, while counter-ECCD leads to negative central shear and formation of an electron ITB with T_e(0)>12 keV. MHD phenomena, especially fishbones, contribute to achieving quasi-stationary advanced discharge conditions and trigger ITBs, which is attributed to poloidal E × B shearing driven by redistribution of resonant fast particles. But MHD instabilities also limit the operational regime of conventional (NTMs) and advanced (double tearing, infernal and external kink modes) scenarios. The onset β_N for NTM is proportional to the normalized gyroradius ρ^*. Complete NTM stabilization was demonstrated at β_N = 2.5 using ECCD at the island position with 10% of the total heating power. MHD limits are expected to be extended using current profile control by off-axis current drive from more tangential NBI combined with ECCD and wall stabilization. Presently, the ASDEX Upgrade divertor is being adapted to optimal performance at higher δ's and tungsten covering of the first wall is being extended on the basis of the positive experience with tungsten on divertor and heat shield tiles.

Research on the Alcator C-Mod tokamak is focused on exploiting compact high density plasmas to understand core transport and heating, the physics of the H mode transport barrier, and the dynamics of the scrape-off layer and divertor. Rapid toroidal acceleration of the plasma core is observed during ohmic heated H modes and indicates a momentum pinch or similar transport mechanism. Core thermal transport observations support a critical gradient interpretation, but with gradients that disagree with present theoretical values. High resolution measurements of the H mode barrier have been obtained, including impurity and neutral densities, and the instability apparently responsible for the favourable `enhanced D alpha' regime has been identified. Divertor bypass dynamic control experiments have directly addressed the important questions surrounding main chamber recycling and the effect of divertor closure on impurities and confinement. Future plans include quasi-steady-state advanced tokamak plasmas using lower hybrid current drive.

Ergodic divertor operation on Tore Supra is characterized by good performance in terms of divertor physics. Control of particle recirculation and impurity screening are related to the symmetry, both poloidally and toroidally, of the shell of open field lines and to its radial extent, Δr ≈ 0.16 m. Feedback control of the divertor plasma temperature has led to controlled radiative divertor experiments. In particular, good performance is obtained when the plasma is controlled to be at a temperature comparable to the energy involved in the atomic processes (15-20 eV). For standard discharges with 5 MW total power and ICRH heating, the low parallel energy flux ≈ 10 MW m^-2 is reduced to ≈ 3 MW m^-2 with nitrogen injection. This is achieved at a modest cost in core dilution, ΔZ_eff ≈ 0.3. Despite the large volume of open field lines ( ≈ 36%), the ergodic divertor does not reduce the possible current in the discharge since stable discharges are achieved with q_sep ≈ 2. It is shown that the reorganization of the current profile in conjunction with a transport barrier in the electron temperature on the separatrix stabilizes the (2,1) tearing mode. Confinement follows the standard L mode confinement. In a few cases at high density and with no gas injection (wall fuelled discharges), `RI-like' modes are reported with a modest increase in confinement ( ≈ 40%). Despite the lack of core fuelling on Tore Supra, high densities during ICRH pulses can be achieved with Greenwald fractions f_G ≈ 1. Compatibility with both ICRH and LH is demonstrated. In particular, long pulse operation with a flat-top in excess of 20 s is achieved with LHCD.

Direct drive laser fusion ignition experiments rely on detailed understanding and control of irradiation uniformity, the Rayleigh-Taylor instability and target fabrication. The Laboratory for Laser Energetics (LLE) is investigating various theoretical aspects of a direct drive National Ignition Facility (NIF) ignition target based on an `all-DT' design: a spherical target of ∼3.4 mm diameter, with a 1-2 μm CH wall thickness and a DT ice layer of ∼340 μm near the triple point of DT (∼19 K). OMEGA experiments are designed to address the critical issues related to direct drive laser fusion and to provide the necessary data to validate the predictive capability of LLE computer codes. The cryogenic targets to be used on OMEGA are hydrodynamically equivalent to those planned for the NIF. The current experimental studies on OMEGA address the essential components of direct drive laser fusion: irradiation uniformity and laser imprinting, Rayleigh-Taylor growth and saturation, compressed core performance and shell-fuel mixing, laser-plasma interactions and their effect on target performance, and cryogenic target fabrication and handling.

MAST is one of the new generation of large, purpose-built spherical tokamaks (STs) now becoming operational, designed to investigate the properties of the ST in large, collisionless plasmas. The first six months of MAST operations have been remarkably successful. Operationally, both merging-compression and the more usual solenoid induction schemes have been demonstrated, the former providing over 400 kA of plasma current with no demand on solenoid flux. Good vacuum conditions and operational conditions, particularly after boronization in trimethylated boron, have provided plasma current of over 1 MA with central plasma temperatures (ohmic) of order 1 keV. The Hugill and Greenwald limits can be exceeded and H mode achieved at modest additional NBI power. Moreover, particle and energy confinement show an immediate increase at the L-H transition, unlike the case of START, where this became apparent only at the highest plasma currents. Halo currents are small, with low toroidal peaking factors, in accordance with theoretical predictions, and there is evidence of a resilience to the major disruption.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current I_p was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 10¹³ cm^-3, increasing the plasma energy to 59 kJ and the toroidal β, β_T, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (P_NBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to β_T ≈ 18% at a plasma current of 1.1 MA.

TJ-II is a four period, low magnetic shear stellarator (R = 1.5 m, a < 0.22 m, B₀ ⩽ 1.2 T) which was designed to have a high degree of magnetic configuration flexibility. In the most recent experimental campaign, coupling of the full ECRH power (P_ECRH ⩽ 600 kW) to the plasma has been possible using two ECRH transmission lines which have different power densities. Both helium and hydrogen fuelled plasmas have been investigated. The article reviews the latest physics results in particle control, configuration effects, and transport and fluctuation studies.

The TCV tokamak (R = 0.88 m, a<0.25 m, B_T<1.54 T) is equipped with six 0.5 MW gyrotron sources operating at 82.7 GHz for second harmonic X mode ECH. By distributing the ECCD current sources over the discharge cross section, fully driven stationary plasmas with I_p = 210 kA, n_e0 = 2 × 10¹⁹m^-3, T_e0 ≈ 4 keV, were obtained for the full discharge duration of 2 s. Highly peaked electron temperature profiles with T_e0 up to 12 keV were obtained in central counter-current drive scenarios with off-axis ECH. Absorption measurements using a 118 GHz gyrotron have demonstrated the importance of suprathermal electrons for third harmonic absorption. A coupled heat-particle transport phenomenon known as `density pumpout', which leads to the expulsion of particles from the plasma core, has been linked to the presence of m = 1 modes, suggesting that it is due to the existence of locally trapped particles associated with the loss of axisymmetry. Highly elongated discharges have been developed with ohmic heating (κ<2.8) and off-axis ECH. The latter exhibit considerably improved vertical stability due to current profile broadening. A `gateway' for ELMy H modes has been discovered, which allows stationary ohmic ELMy H mode operation over a wide range of elongation, triangularity and density. Divertor detachment experiments suggest the existence of recombination pathways other than three body or radiative processes.

Experiments with ECRH performed during 1998-2000 are presented. An operating regime with an edge transport barrier, interpreted as the H mode, has been studied with both on-axis and off-axis ECRH. Experimental data imply formation of an internal transport barrier simultaneously with the external transport barrier in the case of off-axis ECRH. Strong evolution of plasma potential in the regions of the external and internal transport barriers has been observed using heavy ion beam probe diagnostics. The mode of improved confinement with peaked density profiles and increased ion temperature has been observed after pellet injection into ECRH plasmas. Instabilities, identified as neoclassical tearing modes, have been found to limit β in the regimes with a high fraction of bootstrap current. The dependence of the critical β on the q(r) profile has been observed. A systematic study of plasma turbulence has been started using correlation reflectometry diagnostics.

The main purpose of TRIAM-1M (R₀ = 0.8 m, a × b = 0.12 m × 0.18 m, B = 8 T) is to study the route towards a high field compact steady state fusion reactor. In the advanced steady state operation programme, a heating mechanism for the high ion temperature mode with an internal transport barrier has been studied, an enhanced current drive mode in an extended (higher power and higher density) operation regime has been obtained, current density profile control experiments using multicurrent drive systems have been performed and the effects of wall recycling, wall pumping and wall saturation on particle control have been investigated.

Tokamak experimental research in China has made important progress. The main efforts were related to quasi-steady-state operation, LHCD, plasma heating with ICRF, IBW, NBI and ECRH, fuelling with pellets and supersonic molecular beams, and first wall conditioning techniques. Plasma parameters in the experiments were much improved, for example n_e = 8 × 10¹⁹m^-3 and a plasma pulse length of >10 s were achieved. ICRF boronization and conditioning resulted in Z_eff close to unity. Steady state full LH wave current drive has been achieved for more than 3 s. LHCD ramp-up and recharge have also been demonstrated. The best η_CD^exp ≈ 0.5(1+0.085exp(4.8(B_T-1.45)))n_eI_CDR_p/P_LH = 10¹⁹ m^-2 A W^-1. Quasi-steady-state H-mode-like plasmas with a density close to the Greenwald limit were obtained by LHCD, where the energy confinement time was nearly five times longer than in the ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macroturbulence has been extensively carried out experimentally. AC tokamak operation has been successfully demonstrated.

The magnet design of the new ITER-FEAT machine comprises 18 toroidal field (TF) coils, a central solenoid (CS), 6 poloidal field coils and correction coils. A key driver of this new design is the requirement to generate and control plasmas with a relatively high elongation (κ₉₅ = 1.7) and a relatively high triangularity (δ₉₅ = 0.35). This has led to a design where the CS is vertically segmented and self-standing and the TF coils are wedged along their inboard legs. Another important design driver is the requirement to achieve a high operational reliability of the magnets, and this has resulted in several unconventional designs, and in particular the use of conductors supported in radial plates for the winding pack of the TF coils. A key mechanical issue is the cyclic loading of the TF coil cases due to the out-of-plane loads which result from the interaction of the TF coil current and the poloidal field. These loads are resisted by a combination of shear keys and `pre-compression' rings able to provide a centripetal preload at assembly. The fatigue life of the CS conductor jacket is another issue, as it determines the CS performance in terms of the flux generation. Two jacket materials and designs are under study. Since 1993, the ITER magnet R&D programme has been focused on the manufacture and testing of a CS and a TF model coil. During its testing, the CS model coil has successfully achieved all its performance targets in DC and AC operations. The manufacture of the TF model coil is complete. The manufacture of segments of the full scale TF coil case is another important and successful part of this programme and is near completion. New R&D effort is now being initiated to cover specific aspects of the ITER-FEAT design.

The extensive design effort for KSTAR has been focused on two major aspects of the KSTAR project mission - steady-state-operation capability and advanced tokamak physics. The steady state aspect of the mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long pulse current drive and heating systems. The advanced tokamak aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double null divertor and passive stabilizers, internal control coils and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, the vacuum vessel, plasma facing components and power supplies. The new KSTAR experimental facility with cryogenic system and deionized water cooling and main power systems has been designed, and the construction work is under way for completion in 2004.