Table of contents

The m = 2/n = 1 neoclassical tearing mode (NTM) has been observed to strongly degrade confinement and frequently lead to a disruption in high β discharges in DIII-D if allowed to grow to a large size. Stabilization of grown NTMs by the application of a highly localized electron cyclotron current drive (ECCD) at the island location has led to operation at an increased plasma pressure, up to the no-wall kink limit. After the NTM is stabilized by the ECCD, the correct location for the current drive is maintained using information from real-time equilibrium reconstructions which include measurements from the motional Stark effect diagnostic. This same process is used alternatively to prevent the mode from ever growing, leading to performance at the pressure limit in high performance hybrid discharges with β above 4%. Modelling using the modified Rutherford equation shows that the required power is in close agreement with the experimental threshold for the prevention of the 2/1 NTM.

The 1st Joint (Host Laboratory) Experiment on 'joint research using small tokamaks' was carried out using the IPP Prague experimental facility 'CASTOR tokamak'. The main experimental programme was aimed at characterizing the edge plasma in a tokamak by using different advanced diagnostic techniques. It is widely recognized that characterization of phenomena occurring at the plasma edge is essential for understanding the plasma confinement in a tokamak. The edge plasma in small and large scale experiments has many similar features, and the results obtained through detailed measurements in a small flexible device such as CASTOR are in many aspects still relevant to those in large tokamaks. Therefore, it is expected that the results of this joint experiment will have general validity. The radial and poloidal structure of electrostatic turbulence was characterized. The effects of edge biasing were analysed. Radiation fluctuations and profile measurements were performed using fast bolometry. Plasma position measurements were performed using novel Hall sensors.

A development scenario of the tokamak reactor in three stages (i.e. the experimental reactor ITER, a demonstration reactor and a commercial reactor) has been recently discussed. In order to construct a feasible development strategy, it is necessary to evaluate which component of reactor technologies and to what extent should be developed. From the viewpoint of the future electric supplier, we have proposed the conceptual design of a commercial power plant, compact reversed shear tokamak (CREST), and a demonstration power plant, Demo-CREST. On the other hand, the project of the experimental reactor ITER is underway, and its experimental plan and R&D activities are almost completed. Hence, it is most important and reasonable to investigate the demonstration power plant on the track of ITER in order to show a specific development scenario of the tokamak reactor. In this report, we discuss the engineering aspect in the Demo-CREST design and analyse the critical development issues towards an advanced tokamak CREST. The power flow and power plant system for Demo-CREST are investigated for improvement in the thermal efficiency of a single device, and the development goals for each reactor component and for each development step are quantitatively analysed.

A spontaneous shift of diverted plasma position during a discharge was investigated in a helical-axis heliotron device, Heliotron J. The shift of a few cm was observed for discharges with a non-inductive plasma current <3 kA and the plasma stored energy <3 kJ. The time dependence of the observed shift was related to that in the plasma current more closely than that in the stored energy. The most plausible mechanism for the observed shift is the change in the edge field topology caused by the plasma current. The three-dimensional finite-β equilibrium calculations assuming profiles of plasma pressure and plasma current density indicate that the effect of the plasma current on the field topology depends not only on the current flow direction but also on the current density profile. This experiment points out not only the importance of current control in fixing the divertor plasma position in a low-shear helical device but also the possibility of 'divertor swing' for the reduction of the divertor particle/heat load by controlling a small amount of plasma current within a tolerable influence on the plasma performance.

Single-particle orbits in a low aspect ratio axisymmetric configuration have been studied to clarify their role in current initiation using electron cyclotron waves. In this study, electron guiding centre orbits are studied neglecting collisions and charge exchange. Current is generated mainly by the toroidal drift of mirror-trapped electrons. Analytic scaling laws for the current and the poloidal field generation are derived, and they are found to be functions of temperature, poloidal field strength, number of electrons and parameters defining the magnetic and geometrical configuration. The current increases with temperature and decreases with the poloidal field strength as long as the temperature is much smaller than the characteristic temperature, whose thermal velocity is defined by the product of poloidal cyclotron frequency and the major radius.

First principles expressions are given for the parameters governing collisional diffusion and parallel losses of mass, momentum and energy in tokamak scrape-off layer (SOL) plasmas. These dissipative, or damping, coefficients are based on neoclassical perpendicular transport (Pfirsch–Schlüter diffusion) and classical parallel transport (sub-sonic advection and Spitzer–Härm diffusion). When numerical values derived from these expressions are used to compute damping coefficients for the edge-SOL electrostatic (ESEL) turbulence code, simulations correctly reproduce the radial profiles of particle density, n, and electron temperature, T_e, as well as statistical distributions and temporal correlations of particle density and flux density measured in Ohmic and L-mode plasmas on the TCV tokamak. Similarly, preliminary calculations agree reasonably well with radial profiles of T_e measured in Ohmic and L-mode plasmas on JET, although the particle density e-folding length is over-estimated by a factor of 3; this discrepancy is largely removed by reducing the parallel density gradient length by a factor measuring the poloidal asymmetry (ballooning) of filament displacements. These encouraging results suggest that turbulent SOL transport is driven by interchange motions, caused by unfavourable curvature and strong pressure gradients in the edge region, with the level of turbulence being influenced by neoclassical diffusion and parallel losses in the SOL region. Moreover, the curvature drive offers a viable mechanism for the origin of the B × ∇B-independent part of the parallel SOL flow measured on many tokamaks, including JET and TCV, with ESEL simulation predicting a parallel Mach number of ≈ 0.2 in JET Ohmic and L-mode plasmas, in fair agreement with Mach probe measurements.

Experimental observations in a number of tokamaks operated in the hybrid regime revealed that the presence of a rotating neoclassical island substantially reduces the occurrence of sawtooth. This possibility of a rotating mangetic island driving counter currents near the plasma centre in present-day tokamaks is studied. Three mechanisms are investigated with two of them driving sufficient current. They rely on establishing an oscillating parallel electric field by the rotating neoclassical island. First is the excitation of an electrostatic side band through diamagnetic and curvature drifts; second is the excitation of the kinetic Alfvén wave at the plasma centre through the polarization drifts. The third mechanism is the 'prompt' modification of the energetic particle distribution function by the neoclassical island and was found to be relatively weak. The effect of the energetic particles on the counter-current drive due to modification of the energetic particle distribution function on the long (energetic particle slowing-down) time scale has been simulated using a particle simulation code. A firm conclusion has not yet been drawn. Transport simulations indicate that present experimental observations could be the effect of either a negative current drive in the background plasma or the anomalous transport of the energetic particles.

Pellet injection from the inner wall is planned for use in ITER as the primary core fuelling system since gas fuelling is expected to be highly inefficient in burning plasmas. Tests of the inner wall guide tube have shown that 5 mm pellets with up to 300 m s⁻¹ speeds can survive intact and provide the necessary core fuelling rate. Modelling and extrapolation of the inner wall pellet injection experiments from present day's smaller tokamaks leads to the prediction that this method will provide efficient core fuelling beyond the pedestal region. Using pellets for triggering of frequent small edge localized modes is an attractive additional benefit that the pellet injection system can provide. A description of the ITER pellet injection system's capabilities for fuelling and ELM triggering is presented and performance expectations and fusion power control aspects are discussed.

Edge cooling experiments with a tracer-encapsulated solid pellet in the large helical device (LHD) show a significant rise in core electron temperature (the maximum rise is around 1 keV) as well as in many tokamaks. This experimental result indicates the possible presence of the nonlocality of electron heat transport in plasmas where turbulence as a cause of anomalous transport dominates. The nonlocal electron temperature rise in the LHD takes place in almost the same parametric domain (e.g. in a low density) as in the tokamaks. Meanwhile, the experimental results of LHD show some new aspects of nonlocal electron temperature rise, for example the delay in the nonlocal rise of core electron temperature relative to the pellet penetration time increases with the increase both in the collisionality in the core plasma and the electron temperature gradient scale length in the outer region of the plasma.

The Japanese contribution to the construction of the ITER superconducting magnet system is to procure 25% of the toroidal field (TF) conductors, about half of the TF coil winding packs, all of the TF coil structures and all of the central solenoid conductors. The Japan Atomic Energy Agency has started extensive technology development for the preparation of these procurements. These include (1) trial fabrication of the superconductors at industry level and their performance demonstration, (2) manufacturing studies on the TF coil winding pack and structures, (3) full scale trial fabrication of TF coil structures and (4) the establishment of a database on the structural materials. All these activities are performed in close collaboration with Japanese industry involving around 20 major companies. Fruitful results have so far been obtained, although further development of the technology is required in some areas, including the qualification of conductors and the establishment of welding procedures of structures.

Construction of the Korea superconducting tokamak advanced research (KSTAR) tokamak is in its final phase. For the long-pulse KSTAR discharges, the ion cyclotron range of frequencies (ICRF) and neutral beam injection (NBI) heating systems are expected to play important roles through a selective heating of ions and electrons, control of the plasma pressure and current profiles, a core fuelling and beam diagnostics for the KSTAR. In addition, the ICRF system is expected to be used for possible discharge cleaning and assisting in the tokamak startup. In this paper, the recent progress in the development of the ICRF and the NBI heating systems is described. The four-strap ICRF antenna has been successfully tested for a voltage up to 41 kV for a pulse length of 300 s (to 46 kV for 20 s) in a test chamber. A prototype KSTAR NBI system has been developed. At present, the system has successfully produced a 1 MW beam power for 200 s and a 3.5 MW output beam power for 4 s.

The integrated core-pedestal-SOL model is applied to the simulation of a typical DEMO operation. Impurity seeding is used to reduce the power load on the divertor to acceptable levels. The influence on long-pulse operation of impurity seeding with various impurities is investigated. DEMO operation at acceptable peak power loads and long-pulse lengths is demonstrated.

Examination of radial electric field (E_r) profiles in the scrape-off layer (SOL) of ASDEX Upgrade (AUG) and JET revealed large discrepancies between 2D fluid edge modelling and experiment. Experimental profiles of plasma potential (V_p) in the outer (low field) side of the plasma, obtained with reciprocating Langmuir probes, decay radially with electron temperature, T_e, with the −eE_r/∇T_e ratio being > 1.5. In contrast, code simulated E_r are fairly low in most of the SOL (compared with −∇T_e/e). Modelling with kinetic treatment of neutrals and drifts was performed using the SOLPS code for AUG cases and EDGE2D-Nimbus for JET cases.

Mismatches between modelled and experimental E_r may be caused by the recently established tendency for the SOLPS code to underestimate T_e in the divertor of AUG. It was attributed to non-locality of parallel transport of supra-thermal, heat-carrying electrons originating upstream of the divertor, which are usually only weakly collisional and can penetrate, with few collisions, to the target. Ratios −eE_r/∇T_e obtained from the probe measurements in JET are of order 1.6, while in AUG, JT-60U and TCV they are of order 3. Such high values point to the possibility of fast electrons contributing, apart from target heat fluxes, also to the formation of the Debye sheath.

The problem of the underestimation of E_r in the codes must be closely related with the well-known problem of the underestimation of those parts of parallel ion flows in the SOL that are influenced by the toroidal field direction. It was demonstrated earlier that parallel ion flow at the outer midplane is dominated by the ion Pfirsch–Schlüter flow, which in turn is partly driven by the radial electric field. The T_e and E_r discrepancies, as well as discrepancies between simulated and experimental parallel ion flows, raise a question of the validity of fluid codes for the plasma edge modelling and prompt the inclusion of kinetic effects into present-day 2D fluid codes which assume strong collisionality.

First β scan experiments have been made in ASDEX Upgrade in H-mode plasmas with type-I ELMs. Two sets of discharges, performed in different density ranges, scan β between β_N = (1.4–2.2) and β_N = (1.4–2). Global analysis of these dedicated experiments exhibits a strong unfavourable β scaling as Bτ_th ∝ β^−0.9. This tendency is confirmed by a local analysis which shows an increase in the thermal heat diffusivity with increasing β as χ_eff/B ∝ β^0.65. These results are discussed and compared with studies made in DIII-D, JET and JT-60U.