Note: Internal diamagnetic flux measurements on ASDEX Upgrade.

Internal diamagnetic flux measurements, with measurement loops and compensation magnetic probes inside the vacuum vessel, are now available on the ASDEX Upgrade tokamak. The measured diamagnetic flux is compared to that predicted by simulations and calculated from equilibrium reconstruction. The diamagnetic flux measured at 2 positions separated toroidally by 180° in the vacuum vessel is compared.

Real-time diamagnetic flux measurements on ASDEX Upgrade.

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

First observation of edge localized modes mitigation with resonant and nonresonant magnetic perturbations in ASDEX Upgrade.

First experiments with nonaxisymmetric magnetic perturbations, toroidal mode number n=2, produced by newly installed in-vessel saddle coils in the ASDEX Upgrade tokamak show significant reduction of plasma energy loss and peak divertor power load associated with type-I edge localized modes (ELMs) in high-confinement mode plasmas. ELM mitigation is observed above an edge density threshold and is obtained both with magnetic perturbations that are resonant and not resonant with the edge safety factor profile. Compared with unperturbed type-I ELMy reference plasmas, plasmas with mitigated ELMs show similar confinement, similar plasma density, and lower tungsten impurity concentration.

Direct observation of current in type-I edge-localized-mode filaments on the ASDEX Upgrade tokamak.

Magnetically confined plasmas in the high confinement regime are regularly subjected to relaxation oscillations, termed edge localized modes (ELMs), leading to large transport events. Present ELM theories rely on a combined effect of edge current and the edge pressure gradients which result in intermediate mode number (n≅10-15) structures (filaments) localized in the perpendicular plane and extended along the field lines. By detailed localized measurements of the magnetic field perturbation associated to type-I ELM filaments, it is shown that these filaments carry a substantial current.

Convective and diffusive energetic particle losses induced by shear Alfvén waves in the ASDEX upgrade tokamak.

We present here the first phase-space characterization of convective and diffusive energetic particle losses induced by shear Alfvén waves in a magnetically confined fusion plasma. While single toroidal Alfvén eigenmodes (TAE) and Alfvén cascades (AC) eject resonant fast ions in a convective process, an overlapping of AC and TAE spatial structures leads to a large fast-ion diffusion and loss. Diffusive fast-ion losses have been observed with a single TAE above a certain threshold in the fluctuation amplitude.

Soft x-ray virtual diagnostics for tokamak simulations.

The numerical toolset, FAR-TECH Virtual Diagnostic Utility, for generating virtual experimental data based on theoretical models and comparing it with experimental data, has been developed for soft x-ray diagnostics on DIII-D. The virtual (or synthetic) soft x-ray signals for a sample DIII-D discharge are compared with the experimental data. The plasma density and temperature radial profiles needed in the soft x-ray signal modeling are obtained from experimental data, i.e., from Thomson scattering and electron cyclotron emission. The virtual soft x-ray diagnostics for the equilibriums have a good agreement with the experimental data. The virtual diagnostics based on an ideal linear instability also agree reasonably well with the experimental data. The agreements are good enough to justify the methodology presented here for utilizing virtual diagnostics for routine comparison of experimental data. The agreements also motivate further detailed simulations with improved physical models such as the nonideal magnetohydrodynamics contributions (resistivity, viscosity, nonaxisymmetric error fields, etc.) and other nonlinear effects, which can be tested by virtual diagnostics with various stability modeling.

Fast-ion losses due to high-frequency MHD perturbations in the ASDEX upgrade Tokamak.

Time-resolved energy and pitch angle measurements of fast-ion losses correlated in frequency and phase with high-frequency magnetohydrodynamic perturbations have been obtained for the first time in a magnetic fusion device and are presented here. A detailed analysis of fast-ion losses due to toroidal Alfvén eigenmodes has revealed the existence of a new core-localized magnetohydrodynamic perturbation, the sierpes mode. The sierpes mode is a non-Alfvénic instability which dominates the losses of fast ions in ion cyclotron resonance heated discharges, and it is named for its footprint in the spectrograms ("sierpes" means "snake" in Spanish). The sierpes mode has been reconstructed by means of highly resolved multichord soft-x-ray measurements.

Magnetically driven filament probe.

A radially movable probe has been developed for studies of filamentary transport in ASDEX Upgrade during edge localized modes (ELMs) by means of Langmuir tips and magnetic pickup coils. The probe is permanently installed at the low field side in the ASDEX Upgrade vacuum vessel and is not subject to limitations in probe size, as, for example, probes on a shared manipulator are. The probe is moved by a magnetic drive, which allows for easy installation in the vessel, and has moderate machine requirements, as it will only require an electric feedthrough and an external power supply. The drive gives a linear motion with a radial range of 5 cm within 50 ms, where range and velocity can be largely scaled according to experimental requirements. The probe has been installed in the outer midplane of the ASDEX Upgrade vessel, where ELM filaments are expected to have their maximum amplitude. Filaments are coherent substructures within an ELM, carrying a fraction of the ELM released energy towards the wall. The new probe allows to measure the structure of these filaments, in particular, parameters such as filament rotation (by time delay measurements) and size (by peak width analysis). Activating the drive moves the probe from a safe position behind the limiter to a position in front of the limiters, i.e., exposes the Langmuir pins to the scrape-off layer plasma.

Enhancement of the stabilization efficiency of a neoclassical magnetic island by modulated electron cyclotron current drive in the ASDEX upgrade tokamak.

The efficiency of generating a helical current in magnetic islands for the purpose of suppression of neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD) is studied experimentally in the ASDEX Upgrade tokamak. It is found that the efficiency of generating helical current by continuous current drive in a rotating island drops drastically as the width 2d of the co-ECCD driven current becomes larger than the island width W. However, by modulating the co-ECCD in phase with the rotating islands O point, the efficiency can be 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, suggesting that modulation capability should be foreseen.

Reaching high poloidal beta at Greenwald density with internal transport barrier close to full noninductive current drive.

In the ASDEX Upgrade tokamak, high poloidal beta up to beta(pol) = 3 at the Greenwald density with H-mode confinement has been reached. Because of the high beta, the plasma current is driven almost fully noninductively, consisting of 51% bootstrap and 43% neutral beam driven current. To reach these conditions the discharge is operated at low plasma current ( I(P) = 400 kA) and high neutral beam heating power ( P(NBI) = 10 MW). The discharge combines an edge (H mode) and internal transport barrier at high densities without confinement-limiting MHD activities. The extrapolation to higher plasma currents may offer a promising way for an advanced scenario based fusion reactor.

High-confinement regime at high beta(N) values due to a changed behavior of the neoclassical tearing modes.

High confinement [with H = 1, ITERH- 98( y,2)] at beta(N) approximately 2.3 has been found in ASDEX Upgrade discharges with existing (m,n) = (3,2) neoclassical tearing modes (NTMs). The reason for the high confinement at high beta(N) values is a transition of the NTMs into the so-called frequently interrupted regime. In this regime, the NTM growth is frequently interrupted by sudden drops in amplitude. Because of the long NTM growth time, the resulting averaged NTM amplitude remains much smaller than the corresponding saturated value. As this behavior with the beneficial effect on energy confinement has been confirmed at JET, such a high confinement regime at beta(N)>2 might also be expected for ITER.

Simultaneous attainment of high electron and ion temperatures in discharges with internal transport barriers in ASDEX upgrade

Internal transport barriers have been demonstrated to exist also under conditions with T(e) approximately T(i) approximately 10 keV and predominant electron heating of the tokamak core region. Central electron cyclotron heating was added to neutral beam injection-heated ASDEX Upgrade discharges with a preexisting internal transport barrier, established through programmed current ramping leading to shear reversal. Compared to a reference internal transport barrier discharge without electron cyclotron resonance heating, the electron heat conductivity in the barrier region was found not to increase, in spite of a fivefold increase in electron heat flux, and also angular momentum and ion energy transport did not deteriorate.

Complete suppression of neoclassical tearing modes with current drive at the electron-cyclotron-resonance frequency in ASDEX upgrade tokamak

Noninductive current drive has been performed in the tokamak ASDEX upgrade by injection of radiofrequency waves at the second harmonic of the electron-cyclotron frequency in order to suppress unwanted disturbances of the magnetic-field configuration. The current has been driven parallel [co-electron cyclotron current drive (ECCD)] and antiparallel (counter-ECCD) to the plasma current to compare the effect of heating with direct current drive in the magnetic island. For the first time it has been shown experimentally that total stabilization of neoclassical tearing modes is possible with co-ECCD. The experiments verify the role of direct current drive as opposed to local heating.