The electron heat transport in low density H-mode plasmas heated by neutral beam injection (NBI) is investigated in ASDEX Upgrade using electron cyclotron heating (ECH) combining both steady-state and transient response analysis by modulating the ECH power. Under these conditions, more than 60% of the NBI power (>3 MW) is delivered to the ions, while approximately 20% (∼1 MW) is delivered to the electrons. In the confinement region, the electron-to-ion temperature ratio, Te/Ti, varies between 0.5 and 0.7 in the NBI-only phase and between 0.8 and 1.0 when the ECH is also applied. Due to the low collisional coupling, the power in the electron channel is locally more than doubled by applying up to the available 2 MW of ECH, while the power in the ion channel is locally increased by less than 30%. A dependence on the density of the reaction of the plasma parameters to the ECH is observed. For plasmas with average density (defined as 'hot-ion' H-modes), when the ECH is applied, Te increases, the central Ti drops and the density flattens. These effects disappear with increasing density and are not observed for (defined as 'regular' H-modes). Power balance analysis of both the hot-ion and regular H-modes points to a strong resilient behaviour of the Te profiles. In the hot-ion cases, the ECH heating induces a strong increase in transport in the ion channel. Power balance and transient response analysis of the regular H-modes are consistent with an inverse scale length transport model with a threshold in , above which the electron heat transport is increased. Comparison with recent studies in pure EC heated L-modes points to a stronger resilience of Te in the NBI heated H-modes.
A steady-state, fully noninductive plasma current has been sustained for the first time in a tokamak using electron cyclotron current drive only. In this discharge, 123 kA of current have been sustained for the entire gyrotron pulse duration of 2 s. Careful distribution across the plasma minor radius of the power deposited from three 0. 5-MW gyrotrons was essential for reaching steady-state conditions. With central current drive, up to 153 kA of current have been fully replaced transiently for 100 ms. The noninductive scenario is confirmed by the ability to recharge the Ohmic transformer. The dependence of the current drive efficiency on the minor radius is also demonstrated.
An experimental study of the extraordinary mode (X mode) absorption at the third electron cyclotron harmonic frequency has been performed on the TCV tokamak in plasmas preheated by X mode at the second harmonic. Full single pass absorption of injected X3 power was measured with X2 preheating in co-current drive (CO-ECCD). The measured absorption exceeds that predicted by the linear ray tracing code TORAY-GA by more than a factor of 2 for the CO-ECCD case. Experimental evidence indicates that a large fraction of the X3 power is absorbed by electrons in an energetic tail created by the X2 ECCD preheating.
This paper reports on the experimental studies, performed in the tokamak ASDEX Upgrade, for increasing the efficiency in generating a helical current within magnetic islands with the purpose of suppressing neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD). It is shown that the efficiency of generating this current by continuous CD in a rotating island drops drastically as the width 2d of the co‐ECCD driven current becomes larger than the islands size W. However, by modulating the co‐ECCD in phase with the island's O‐point, the efficiency is recovered. The results are in good agreement with theoretical calculations taking into account the equilibration of the externally driven current on the island flux surfaces. The result is especially important for large next‐step fusion devices, such as ITER, where 2d>W is expected to be unavoidable during NTM suppression, indicating that modulation capability should be foreseen.
The ECCD deposition width during NTM stabilisation experiments has been scanned over a wide range relative to the island width. The highest efficiency in terms of stabilisation and achievable β N with suppressed NTM could be achieved with a narrow deposition with the highest current density IECCD/d. For a broad deposition no stabilisation could be achieved. However, the NTM could be stabilised with modulated broad ECCD, depositing only in the islands O-point. Introduction Neoclassical Tearing Modes (NTMs) are of great concern for tokamak plasmas, in particular for ITER and a future fusion reactor, as the presence of a (3/2)-NTM reduces the achievable βN by at least 10-20%. As the achievable fusion power is proportional to β 2 this is not tolerable. Schemes for the stabilisation of NTMs by local co - Electron Cyclotron Current Drive (co-ECCD) with respect to the plasma current have been developed at various experiments (1-3) and established as a tool with feedback capabilities. Therefore in ITER a system with 24 MW ECCD power at 170 GHz has been foreseen for this purpose. Most of present experiments are performed with a fixed ECCD deposition width d, which is smaller than the marginal island size Wmarg. For W < Wmarg an NTM naturally decays away and βN can rise again independently of the NTM size. For ITER, the condition d < Wmarg is not necessarily fulfilled, and possible ways to reliably stabilise an NTM have to be developed for this purpose. The effect of the ratio d/Wmarg has been investigated. The mod- ulation of the ECCD depositing only power in the islands O-point is predicted to improve the effi- ciency of the stabilisation. This has been proved experimentally and is compared with predictions.
The “improved H-mode,” realized in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] in 1998, demonstrates that advanced requirements beyond the standard H-mode for confinement [confinement enhancement factor H98(y,2)>1], stability (normalized beta βN∼3–3.5) and, at densities close to Greenwald density, exhaust can be simultaneously met and maintained stationary for several resistive diffusion times. The q profile is characterized by low central magnetic shear and axis safety factor q0>1 that is obtained by particular heating and current ramp-up scenarios and maintained via benign instabilities. Core transport is still governed by drift-wave turbulence with stiff temperature profiles, but density profiles are more strongly peaked and contribute to the increase in confinement. Neoclassical tearing modes remain small, enabling routine operation up to βN∼3 at international thermonuclear experimental reactor (ITER) relevant collisionalities, for normalized Lamor radii down to four times the ITER value and for a broad range of q95=3.2–4.5. Using tailored heat deposition including central wave heating a compromise was found in density peaking for enhanced confinement and limiting the high-Z impurity concentrations even with a tungsten-coated first wall and divertor. As far as the ITER [ITER EDA Documentation Series No. 24, 2002] relevance of this regime is concerned, its compatibility with significant central electron heating, high edge densities, and type-II edge localized modes is of importance. The GLF23 turbulence model predicts still peaked density profiles and sufficient transport to avoid impurity accumulation. The fusion performance in terms of βNH98(y,2)∕q952 is nearly doubled compared with the ITER base-line scenario at low-q values, while at medium q’s bootstrap current fractions up to 50% and long inductive pulse lengths allow ITER “hybrid” operation.
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