S. Bernabei

Dynamical modelling of tearing mode stabilization by RF current drive

The theory of tearing mode stabilization in toroidal plasmas by RF driven currents that are modulated in phase with the island rotation is investigated. A time-scale analysis of the phenomena involved indicates that transient effects, such as finite time response of the driven currents, island rotation during the power pulses and the inductive response of the plasma, are intrinsically important. A dynamical model of such effects is developed, based on a 3-D Fokker-Planck code coupled to both the electric field diffusion equation and the island evolution equation. Extensive applications to both ECCD and LHCD in ITER are presented.

Angular distribution of the bremsstrahlung emission during lower hybrid current drive on PLT

The bremsstrahlung emission from the PLT tokamak during lower-hybrid current drive has been measured as a function of angle between the magnetic field and the emission direction. The emission is peaked strongly in the forward direction, indicating a strong anisotropy of the electron velocity distribution. The data demonstrate the existence of a nearly flat tail of the velocity distribution, which extends out to approximately 500 keV and which is interpreted as the plateau created by Landau damping of the lower-hybrid waves.

Wave-particle studies in the ion cyclotron and lower hybrid ranges of frequencies in alcator C-mod

Abstract This paper reviews the physics and technology of wave-particle-interaction experiments in the ion cyclotron range of frequencies (ICRF) and the lower hybrid (LH) range of frequencies (LHRF) on the Alcator C-Mod tokamak. Operation of fixed frequency (80 MHz) and tunable (40- to 80-MHz) ICRF transmitters and the associated transmission system is described. Key fabrication issues that were solved in order to operate a four-strap ICRF antenna in the compact environment of C-Mod are discussed in some detail. ICRF heating experiments utilizing the hydrogen (H) and helium-3 (3He) minority heating schemes are described, and data are presented demonstrating an overall heating efficiency of 70 to 90% for the (H) minority scheme and somewhat lower efficiency for (3He) minority heating. Mode conversion electron heating experiments in D(3He), D(H), and H(3He) discharges are also reported as well as simulations of these experiments using an advanced ICRF full-wave solver. Measurements of mode-converted ion cyclotron waves and ion Bernstein waves using a phase contrast imaging diagnostic are presented and compared with the predictions of a synthetic diagnostic code that utilizes wave electric fields from a full-wave solver. The physics basis of the LH current profile control program on Alcator C-Mod is also presented. Computer simulations using a two-dimensional (velocity space) Fokker Planck solver indicate that ~200 kA of LH current can be driven in low-density H-mode discharges on C-Mod with ~3 MW of LHRF power. It is shown that this off-axis LH current drive can be used to create discharges with nonmonotonic profiles of the current density and reversed shear. An advanced tokamak operating regime near the ideal no-wall β limit is described for C-Mod, where ~70% of the current is driven through the bootstrap effect. The LH power is coupled to C-Mod through a waveguide launcher consisting of four rows (vertically) with 24 guides per row (toroidally). A detailed description of the LH launcher fabrication is given in this paper along with initial operation results.

Modelling of the electron distribution based on bremsstrahlung emission during lower-hybrid current drive on PLT

Lower-hybrid current drive requires the generation of a high-energy electron tail anisotropic in velocity. Measurements of bremsstrahlung emission produced by this tail are compared with the calculated emission from reasonable model distributions. The physical basis and the sensitivity of this modelling process are described, and the plasma properties of current-driven discharges which can be derived from the model are discussed.

Lower hybrid experiments on PLT using grills with various n∥ spectral widths

Coupling structures for lower hybrid current drive experiments have, until now, been smaller than a free space wavelength and have had a correspondingly broad wavenumber spectrum. The paper reports the results of experiments on the PLT tokamak using a 16-waveguide grill (2.2 wavelengths) which produces a very narrow n∥ = k∥c/ω spectrum. Experimental results from the 16-waveguide grill are compared with results from three other PLT grills with less sharply defined n1 spectra. The current drive figure of merit, , is approximately 40% higher for the experiments with the 16-waveguide coupler than for previously reported experiments on PLT, in spite of the larger spectral gap. The experimental results are consistent with the first-pass damping of a large fraction of the launched spectrum.

Overview of the Alcator C-Mod program

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

Ion cyclotron range of frequencies heating and flow generation in deuterium–tritium plasmas

Recent radio-frequency heating experiments on the Tokamak Fusion Test Reactor (TFTR) [Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] have focused on developing tools for both pressure and current profile control in deuterium–tritium (DT) plasmas. A new antenna was added to investigate pressure profile control utilizing direct ion Bernstein wave (IBW) heating. This was the first time direct IBW heating was explored on TFTR. Plasma heating and driven poloidal flows are observed. Previously heating and current drive via mode-converted IBW waves had been demonstrated in non-DT plasmas but efforts in DT plasmas had been unsuccessful. This lack of success had been ascribed to the presence of a small 7Li minority ion population. In the most recent experiments 6Li was used exclusively for machine conditioning and mode-conversion heating consistent with theory is now observed in DT plasmas.

Overview of the Alcator C-MOD Research Program

Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower hybrid current drive (LHCD) experiments (PLH < 900?kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at Ip ~ 1.0?MA with good efficiency, Idrive = 0.4 PLH/neoR (MA, MW, 1020?m?3,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (Bth/B ? R0.68?0.19) is inferred from its measured variation with BT that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfv?n cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as , and show a significant dependence on X-point topology. Fast camera images of intermittent turbulent structures at the plasma edge show they travel coherently through the SOL with a broad radial velocity distribution having a peak at about 1% of the ion sound speed, in qualitative agreement with theoretical models. Fast D? diagnostics during gas puff imaging show a complex behaviour of discrete ELMs, starting with an n ? 10 precursor oscillation followed by a rapid primary ejection as the pedestal crashes and then multiple, slower secondary ejections.

The KSTAR tokamak

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.

Modelling of advanced tokamak scenarios with LHCD in Alcator C-Mod

A combined model for current profile control and MHD stability analysis has been used to identify stable operating modes near the ideal stability limit (?N 3) in the Alcator C-Mod tokamak. These discharges are characterized by relatively high fractions of bootstrap current (fBS = 0.70) and non-monotonic profiles of the safety factor with qmin > 2. In the absence of a conducting shell, stability was determined by the onset of the low (n = 1) external kink mode. In these studies, current profile control in the plasma periphery (r/a 0.5) was provided by 2.5-3.0?MW of LHCD power. Internal and edge transport barriers were introduced into the model calculations in the form of density transitions. Excellent wave accessibility and absorption were still found in the presence of an H-mode-like edge density barrier. However, the presence of these barriers resulted in about a 10% decrease in the stability limit, from ?N 3 to ?N 2.7.