A5017352704- Magnetic confinement fusion research
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Max Planck Institute for Plasma Physics
2016-2025
Jožef Stefan Institute
2022
Max Planck Society
2012-2021
Polytechnic University of Turin
2021
Ludwig-Maximilians-Universität München
2009-2020
Princeton Plasma Physics Laboratory
1997-2019
Max Planck Innovation
1993-2017
École Polytechnique Fédérale de Lausanne
2007-2011
Karlsruhe Institute of Technology
1995-2010
CEA Cadarache
2008
Progress in the area of MHD stability and disruptions, since publication 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137–2664), is reviewed. Recent theoretical experimental research has made important advances both understanding control tokamak plasmas. Sawteeth are anticipated baseline ELMy H-mode scenario, but tools exist to avoid or them through localized current drive fast ion generation. Active other instabilities will most likely be also required ITER. Extrapolation from...
The phenomenology of edge localized modes (ELMs), an MHD instability occurring in the H-mode plasmas toroidal magnetic fusion experiments, is described. ELMs are important to obtain experimental control particle inventory plasmas. From analysis ELM behaviour different three distinct types identified, namely dithering cycles, type III and I ELMs. A physical picture these phenomena established on grounds theoretical models put forward describe phenomena.
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 type-I edge localized modes (ELMs) high-confinement plasmas. ELM mitigation is observed above an density threshold obtained both perturbations that are resonant not safety factor profile. Compared unperturbed ELMy reference plasmas,...
The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed short pulse and predicted by the ideal MHD theory. Recent experiments, particular those simulating International Thermonuclear Experimental Reactor (ITER) [M. Rosenbluth et al., Plasma Physics Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1995), Vol. 2, p. 517] scenarios with νe*, are often limited low-m/n nonideal...
The physics base for the ITER Physics Design Guidelines is reviewed in view of application to DEMO and areas are pointed out which improvement needed arrive at a consistent set Guidelines. Amongst proposed improvements, area power exhaust plays crucial role since predictive capability present-day models low this expected play major limiting designs due much larger value Ptot/R than devices even ITER.
Noninductive current drive has been performed in the tokamak ASDEX upgrade by injection of radiofrequency waves at second harmonic electron-cyclotron frequency order to suppress unwanted disturbances magnetic-field configuration. The driven parallel [co-electron cyclotron (ECCD)] and antiparallel (counter-ECCD) plasma compare effect heating with direct magnetic island. For first time it shown experimentally that total stabilization neoclassical tearing modes is possible co-ECCD. experiments...
The reduction of neoclassical tearing modes by ECCD is demonstrated experimentally. It shown that with an averaged power only 4-8% the total heating injected into discharge, island width can be reduced 40%, provided centre deposition very close to resonant surface. in mode amplitude results a partial recovery loss stored energy induced mode. This experimental result well reproduced modelling calculations.
Local edge parameters on the ASDEX Upgrade tokamak are investigated at L-mode to H-mode transition, during phases with various types of edge-localized modes (ELMs), and density limit. A scaling law for boundary electron temperature, , is found which describes threshold deuterium-puffed discharges favourable ion -drift direction. The region stable operation bounded by type I ELMs near ideal ballooning limit a minimum temperature necessary avoid thermal instability plasma edge. Stationary III...
Neoclassical tearing modes (NTMs) will be the principal limit on performance in ITER standard scenario, which has beta well below ideal kink limit. Measurements of island size from ASDEX Upgrade, DIII-D and JET rampdown experiments are used to determine marginal for m/n = 3/2 NTM removal. This is compared with data JT-60U removal by electron cyclotron current drive (ECCD) at near constant beta. The empirical consistent both sets found about twice ion banana width. A common methodology...
In the European fusion roadmap, ITER is followed by a demonstration power reactor (DEMO), for which conceptual design under development. This paper reports first results of coherent effort to develop relevant physics knowledge that (DEMO Physics Basis), carried out experts. The program currently includes investigations in areas scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.
Abstract A large scale program to develop a conceptual design for demonstration fusion power plant (DEMO) has been initiated in Europe. Central elements are the baseline points, which developed by system codes. The assessment of credibility these points is often hampered missing information. main physics and technology content central European codes have published (Kovari et al 2014 Fusion Eng. Des . 89 3054–69, 2016 104 9–20, Reux 2015 Nucl. 55 073011). In addition, this publication...
Abstract The ASDEX Upgrade (AUG) programme is directed towards physics input to critical elements of the ITER design and preparation operation, as well addressing issues for a future DEMO design. Since 2015, AUG equipped with new pair 3-strap ICRF antennas, which were designed reduction tungsten release during operation. As predicted, factor two on ICRF-induced W plasma content could be achieved by sheath voltage at antenna limiters via compensation image currents central side straps in...
The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical physics limits, while its front-end components be subject strong adverse effects within the nuclear high temperature environment. ongoing developments ITER D&C represent an important starting point progressing towards DEMO. Requirements detailed exploration of are however pushing design using sophisticated methods aiming large spatial...
For several reasons the challenge to keep loads first wall within engineering limits is substantially higher in DEMO compared ITER. Therefore pre-conceptual design development for that currently ongoing Europe needs be based on load estimates are derived employing most recent plasma edge physics knowledge.
This paper is part of a series publications concerning the development European DEMO during Pre-Concept Design Phase (2014-2020), and also describing strategy for next phase. In particular, it deals with physics basis plasma scenarios employed definition various baselines released so far, assumptions adopted where necessary. course Phase, some these have been progressively replaced results dedicated modelling activities or code developments in general, which are summarized here. The...
H modes with good confinement and small ELMs the characteristics of type II or grassy have been observed on ASDEX Upgrade. Such an ELM behaviour is essential to minimize erosion divertor tiles in any next step device. For first time, operation this favourable could be demonstrated close Greenwald density. Even for such high densities, energy times were recent mode scalings. High density even seems favourable, since steady state pure ELMy phases Upgrade are obtained only above e/GW ⩾ 0.85....
Feedback-controlled puffing of neon and deuterium has been applied to control the edge-localized-mode behavior target plate power deposition during high-power $H$-mode discharges in ASDEX Upgrade. A regime found which more than 90% heating is lost through radiation divertor detachment occurs, without deterioration energy confinement. The plasma remains $H$ mode, exhibiting small-amplitude, high-frequency ELM's, do not penetrate plates strike zone region.
The generalized Rutherford equation for the neoclassical tearing mode is studied. New analytical expressions nonlinear stability criterion, seed island width, and saturated width are derived. These especially useful when small. A formalism calculating current needed to stabilize established by adding an externally driven current. Inserting reference parameters of International Thermonuclear Experimental Reactor (ITER) [ITER-JCT Home Teams, Plasma Phys. Controlled Fusion 37, A19 (1995)], a...
A scintillator based detector for fast-ion losses has been designed and installed on the ASDEX upgrade (AUG) tokamak [A. Herrmann O. Gruber, Fusion Sci. Technol. 44, 569 (2003)]. The resolves in time energy pitch angle of induced by magnetohydrodynamics (MHD) fluctuations. use a novel material with very short decay high quantum efficiency allows to identify MHD fluctuations responsible ion through Fourier analysis. Faraday cup (secondary plate) embedded behind plate an absolute calibration...
The efficiency of generating a helical current in magnetic islands for the purpose suppression neoclassical tearing modes (NTMs) by electron cyclotron drive (ECCD) is studied experimentally ASDEX Upgrade tokamak. It found that continuous rotating island drops drastically as width 2d co-ECCD driven becomes larger than W. However, modulating phase with O point, can be recovered. results are good agreement theoretical calculations taking into account equilibration externally on flux surfaces....
Abstract The achievable efficiency for external current drive through electron-cyclotron waves in a demonstration tokamak reactor is investigated. Two possible designs, one steady state and pulsed operation, are considered. Beam propagation, absorption modelled employing the beam-tracing technique including momentum conservation electron–electron collisions. It found that midplane injection limited by second-harmonic at levels consistent with previous studies. Higher efficiencies can be...
The ITER electron cyclotron (EC) upper port antenna (or launcher) is nearing completion of the detailed design stage and final build-to-print will soon start. main objective this launcher to drive current locally stabilize neoclassical tearing modes (NTMs) (depositing ECCD inside island that forms on either q = 3/2 or 2 rational magnetic flux surfaces) control sawtooth instability (deposit near 1 surface). should be capable steering focused beam deposition location resonant surface over...
Neoclassical tearing modes (NTMs) are magnetic islands which increase locally the radial transport and therefore degrade plasma performance. They self-sustained by bootstrap current perturbed enhanced transport. The confinement degradation is proportional to island width position of resonant surface. q = 2 NTMs much more detrimental than 3/2 due their larger radii. metastable in typical scenarios with βN ⩾ 1 region where safety factor increasing radius. This fact that local pressure gradient...