Table of contents

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semiclosed pumped divertor with a W shaped geometry. Plasma performance in transient and quasi-steady states has been significantly improved in reversed shear and high β_p regimes. The reversed shear regime elevated an equivalent Q^eq_DT transiently up to 1.25 (n_D(0) τ_ET_i(0) = 8.6 × 10²⁰m^-3·s·keV) in a reactor relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high β_p regime with high triangularity, achieving a long sustainment of plasma conditions equivalent to Q^eq_DT ≈ 0.16 (n_D(0) τ_ET_i(0) ≈ 1.4  ×  10²⁰ m^-3·s·keV) for ∼4.5 s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around the ITB, resulting in a strong E_r shear in the ITB region. The W shaped divertor is effective for helium ash exhaust, demonstrating a steady exhaust capability of τ*_He/τ_E ≈ 3-10, which is favourable for ITER. Suppression of neutral backflow and chemical sputtering effects has been observed, while MARFE onset density is rather decreased. Negative ion based neutral beam injection (N-NBI) experiments have created a clear H mode transition. An enhanced ionization cross-section due to multistep ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in good agreement with the theoretical prediction. N-NBI induced TAEs characterized as persistent and bursting oscillations have been observed from a low hot β of ⟨β_h⟩ ≈ 0.1-0.2% without a significant loss of fast ions.

The Joint European Torus (JET) has recently operated with deuterium-tritium (DT) mixtures, carried out an International Thermonuclear Experimental Reactor (ITER) physics campaign in hydrogen, deuterium, DT and tritium, installed the Mark IIGB `Gas Box' divertor fully by remote handling and started physics experiments with this more closed divertor. The DT experiments set records for fusion power (16.1 MW), ratio of fusion power to plasma input power (0.62, and 0.95 ± 0.17 if a similar plasma could be obtained in steady state) and fusion duration (4 MW for 4 s). A large scale tritium supply and processing plant, the first of its kind, allowed the repeated use of the 20 g of tritium on-site to supply 99.3 g of tritium to the machine. The H mode threshold power is significantly lower in DT, but the global energy confinement time is practically unchanged (no isotope effect). Dimensionless scaling `wind tunnel' experiments in DT extrapolate to ignition with ITER parameters. The scaling is close to gyro-Bohm, but the mass dependence is not correct. Separating the thermal plasma energy into core and pedestal contributions could resolve this discrepancy (leading to proper gyro-Bohm scaling for the core) and also account for confinement degradation at high density and at high radiated power. Several radiofrequency heating schemes have been tested successfully in DT, showing good agreement with calculations. Alpha particle heating has been clearly observed and is consistent with classical expectations. Internal transport barriers have been established in optimized magnetic shear discharges in DT and steady state conditions have been approached with simultaneous internal and edge transport barriers. First results with the newly installed Mark IIGB divertor show that the in-out symmetry of the divertor plasma can be modified using differential gas fuelling, that optimized shear discharges can be produced and that krypton gas puffing is effective in restoring L mode edge conditions and establishing an internal transport barrier in such discharges.

The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns were carried out in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 10¹⁹m^-3, central temperatures ranging from 1.4 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scaling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

An overview of TRIAM-1M experiments is presented. The current status of issues related to steady state operation is presented with reference to the achievement of super-ultra-long tokamak discharges sustained by LHCD for over 2 h. The importance of control of the initial phase of the plasma, the avoidance of high heat load concentration, wall conditioning and the avoidance of abrupt termination of long discharges are discussed as the crucial issues for the achievement of steady state operation of the tokamak. A high ion temperature (HIT) discharge fully sustained by 2.45 GHz LHCD with both high ion temperature and steep temperature gradient was successfully demonstrated for longer than 1 min in the limiter configuration. The HIT discharges can be obtained in a narrow window of density and position. The avoidance of heat load concentration on a limiter is the key point for the achievement and long sustainment of the HIT discharge. As the effective thermal insulation between the wall and the plasma is improved for the single null configuration, HIT discharges with peak ion temperature > 5 keV and a steeper temperature gradient of up to 85 keV/m can be achieved through the fine control of density and position. Plasmas with high κ ≈ 1.5 can also be demonstrated for longer than 1 min. The current profile is also well controlled for a time about 2 orders of magnitude longer than the current diffusion time using combined LHCD. The serious damage to the material of the first wall caused by energetic neutral particles produced by charge exchange is also described. As the neutral particles cannot be affected by a magnetic field, this damage by neutral particles must be prevented by a new technique.

The spherical tokamak (ST) is the low aspect ratio limit of the conventional tokamak and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration and was operational at Culham from January 1991 to March 1998, obtaining plasma currents of up to 300 kA and pulse durations of ∼50 ms. Its successor, MAST, is nearing completion and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents of up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high performance central solenoid, with 1.5 MW ECRH and 5 MW of NBI. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose built STs (such as MAST) are described.

Three hohlraum concepts are being pursued at Sandia National Laboratories (SNL) to investigate the possibility of using pulsed power driven magnetic implosions (Z pinches) to drive targets capable of fusion yields in the range 200-1000 MJ. This research is being conducted on SNL's Z facility, which is capable of driving peak currents of 20 MA in various Z pinch load configurations that produce implosion velocities as high as 7.5  ×  10⁷cm/s, X ray energies of 1-2 MJ and X ray powers of 100-250 TW. The first concept, denoted dynamic hohlraum, has achieved a temperature of 180 ± 14 eV in a configuration suitable for driving capsules. In addition, this concept has also achieved a temperature of 230 ± 18 eV in an arrangement suitable for driving an external hohlraum. The second concept, denoted static walled hohlraum, has achieved temperatures of ∼80-100 eV. Experimental investigation of the third concept, denoted Z pinch driven hohlraum, has recently begun. The article discusses each of these hohlraum concepts and provides an overview of the experiments that have been conducted on these systems to date.

The article summarizes six years of technical work carried out by the ITER Joint Central Team and Home Teams under the terms of the ITER Engineering Design Activities (EDA) Agreement. The major products of this effort are as follows: a comprehensive and detailed engineering design with supporting assessments, cost estimates and a schedule based on industrial studies; a comprehensive safety and environmental assessment; and technology R&D to validate and qualify the design, including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information to serve as a basis for a decision on construction. The operation of ITER will demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes.

The article reviews the physics of fusion alpha particles and energetic neutral beam ions studied in the final phase of TFTR operation, with an emphasis on observations in reversed magnetic shear (RS) and enhanced reversed shear (ERS) DT plasmas. Energy resolved measurements of the radial profiles of confined, trapped alphas in RS plasmas exhibit reduced core alpha density with increasing alpha energy, in contrast to plasmas with normal monotonic shear. The measured profiles are consistent with predictions of increased alpha loss due to stochastic ripple diffusion and increased first orbit loss in RS plasmas. In experiments in which a short tritium beam pulse is injected into a deuterium RS plasma, the measured DT neutron emission is lower than standard predictions assuming first orbit loss and stochastic ripple diffusion of the beam ions. A microwave reflectometer measured the spatial localization of low toroidal mode number (n), alpha driven toroidal Alfvén eigenmodes (TAEs) in DT RS discharges. Although the observed ballooning character of the n = 4 mode is consistent with predictions of a kinetic MHD stability code, the observed antiballooning nature of the n = 2 mode is not. Furthermore, the modelling does not show the observed strong dependence of mode frequency on n. These alpha driven TAEs do not cause measurable alpha loss in TFTR. Other Alfvén frequency modes with n = 2-4 seen in both DT and DD ERS and RS discharges are localized to the weak magnetic shear region near q_min. In 10-20% of DT discharges, normal low n MHD activity causes alpha loss at levels above the first orbit loss rate.

The closed ASDEX Upgrade Divertor II, `LYRA', is capable of handling heating powers of up to 20 MW or P/R of 12 MW/m, owing to a reduction of the maximum heat flux to the target plates by more than a factor of 2 compared with the open Divertor I. This reduction is caused by high radiative losses from carbon and hydrogen inside the divertor region and is in agreement with B2-EIRENE modelling predictions. At medium densities in the H mode, the type I ELM behaviour shows no dependence on the heating method (NBI, ICRH). ASDEX Upgrade-JET dimensionless identity experiments showed compatibility of the L-H transition with core physics constraints, while in the H mode confinement, inconsistencies with the invariance principle were established. At high densities close to the Greenwald density, the MHD limited edge pressures, the influence of divertor detachment on separatrix parameters and increasing edge transport lead to limited edge densities and finally to temperatures below the critical edge temperatures for H mode. This results in a drastic increase of the H mode threshold power and an upper H mode density limit with gas puff refuelling. The H mode confinement degradation approaching this density limit is caused by the ballooning mode limited edge pressures and `stiff' temperature profiles relating core and edge temperatures. Repetitive high field side pellet injection allows for H mode operation well above the Greenwald density; moreover, higher confinement than with gas fuelling is found up to the highest densities. Neoclassical tearing modes limit the achievable β depending on the collisionality at the resonant surface. In agreement with the polarization current model, the onset β is found to be proportional to the ion gyroradius in the collisionless regime, while higher collisionalities are stabilizing. The fractional energy loss connected with saturated modes at high pressures is about 25%. A reduction of neoclassical mode amplitude and an increase of β have been demonstrated by using phased ECRH and ECCD in the O point of islands. Advanced tokamak operation with internal transport barriers for both ions and electrons has been achieved with flat shear profiles and q₀ > 1 or with reversed shear and q_min > 2. With flat shear a stationary H mode scenario was maintained for 40 confinement times and several internal skin times with β_N = 2 and H_ITERL-89P = 2.4, where fishbones keep q₀ at 1. β_N is limited by either neoclassical tearing modes in the case of flat shear or kink modes with reversed shear.

Variation of the plasma position relative to the centre of the helical coil winding is a very effective means of controlling the MHD stability and the trapped particle confinement in heliotron/torsatron systems, but improving one of these two characteristics with this parameter simultaneously has a detrimental effect on the other. The inward shifted configuration is favourable for drift orbit optimization but is predicted to be unstable according to the Mercier criterion. Various physics problems, such as electric field structure, plasma rotation and MHD phenomena, have been studied in the Compact Helical System (CHS) with a compromise intermediate position. With this standard configuration, CHS has yielded experimental results that contribute to the understanding of general toroidal confinement physics and low aspect ratio helical systems. In the recent experiments, it was found that a wide range of inward shifted configurations give stable plasma discharges without any restriction to the special pressure profile. Such an enhanced range of operation made it possible to study experimentally the drift orbit optimized configuration in heliotron/torsatron systems. The effect of configuration improvement was studied with plasmas in a low collisionality regime.