Numerical study of tearing mode seeding in tokamak X-point plasma

A detailed understanding of island seeding is crucial to avoid (N)TMs and their negative consequences like confinement degradation and disruptions. In the present work, we investigate the growth of 2/1 islands in response to magnetic perturbations. Although we use externally applied perturbations produced by resonant magnetic perturbation (RMP) coils for this study, results are directly transferable to island seeding by other MHD instabilities creating a resonant magnetic field component at the rational surface. Experimental results for 2/1 island penetration from ASDEX Upgrade are presented extending previous studies. Simulations are based on an ASDEX Upgrade L-mode discharge with low collisionality and active RMP coils. Our numerical studies are performed with the 3D, two fluid, non-linear MHD code JOREK. All three phases of mode seeding observed in the experiment are also seen in the simulations: first a weak response phase characterized by large perpendicular electron flow velocities followed by a fast growth of the magnetic island size accompanied by a reduction of the perpendicular electron velocity, and finally the saturation to a fully formed island state with perpendicular electron velocity close to zero. Thresholds for mode penetration are observed in the plasma rotation as well as in the RMP coil current. A hysteresis of the island size and electron perpendicular velocity is observed between the ramping up and down of the RMP amplitude consistent with an analytically predicted bifurcation. The transition from dominant kink/bending to tearing parity during the penetration is investigated

Multi-device study of temporal characteristics of magnetohydrodynamic modes initiating disruptions

Disruptions in tokamaks are often preceded by magnetohydrodynamic (MHD) instabilities that can rotate or become locked to the wall. Measurements from magnetic diagnostics in the presence of MHD mode precursors to disruptions can yield potentially valuable input to the plasma control system, with a view to disruption avoidance, prediction and mitigation. This paper presents an exploratory analysis of the growth of MHD modes and corresponding time scales on the basis of magnetic measurements in multiple tokamaks. To this end, a database was compiled using disruptive discharges from COMPASS, ASDEX Upgrade, DIII-D and JET, manually classified according to disruption root cause, and characterized by a great diversity of operational conditions and mode dynamics. The typical time during which a mode can be detected using saddle coils and the duration of the locked mode phase in the database both extend over several orders of magnitude, but generally the time scales increase with plasma size. Several additional factors are discussed that can influence these durations, including the disruption root cause. A scaling law for the locked phase duration was estimated, yielding predictions toward ITER of the order of hundreds of milliseconds or even seconds. In addition, a scaling law for the mode amplitude at the disruption onset, proposed earlier by de Vries et al. (2016), is applied to the database, and its predictive capabilities are assessed. Despite significant uncertainty on the predictions from both scaling laws, encouraging trends are observed of the fraction of disruptions that may be detected with sufficient warning time to allow mitigation or even avoidance, based solely on observations of MHD mode dynamics. When combined with similar analysis of measurements from diagnostics that are sensitive to other disruption precursors, our analysis methods and results may contribute to the reliability, robustness and generalization of disruption warning schemes for ITER

Interface and requirements analysis on the DEMO Heating and Current Drive system using Systems Engineering methodologies

In this paper we present the methodology implemented for analyzing System Requirements and Interfaces of the Heating and Current Drive (HCD) system of the European Demonstration Fusion Power Reactor DEMO. The work consisted in updating the preliminary framework of the Model-Based Systems Engineering model of the HCD System Architecture. This is now containing an ontology, a set of 6 perspectives and a defined set of viewpoints for each Perspective, for refining the HCD System Architecture. The scope of the work is to manage the interdependencies of HCD system elements and their integration into DEMO, for a given set of system functions. On the one hand, this means to address the identification and definition of the interfaces occurring, both internally in the HCD system, and between the HCD system and neighboring systems. On the other hand, this implies studying the impact of requirements coming from the ongoing physics studies. The rationale is to provide the technical foreground for supporting the decision-making processes related to the HCD system which is planned to be carried out during the forthcoming Conceptual Design Phase. The results we show in this paper are part of the design and integration activities consisting of both systems engineering methodologies and design analysis, all aiming at ensuring consistency in the overall EU DEMO plant design. In this framework the DEMO Heating and Current Drive system has been selected as pilot project for the application of Systems Engineering methodologies

Completion of the 8 MW Multi-Frequency ECRH System at ASDEX Upgrade

Over the last 15 years, the Electron Cyclotron Resonance Heating (ECRH) system at the ASDEX Upgrade tokamak has been upgraded from a 2 MW, 2 s, 140 GHz system to an 8 MW, 10 s, dual frequency system (105/140 GHz). Eight gyrotrons were in routine operation during the current experimental campaign. All gyrotrons are step-tunable operating at 105 and 140 GHz with a maximum output power of about 1 MW and 10 s pulse length. The system includes 8 transmission lines, mainly consisting of oversized corrugated waveguides (I.D. = 87 mm) with overall lengths between 50 and 70 meters including quasi-optical sections at both ends. Further improvements of the transmission lines with respect to power handling and reliability are underway

Exploring fusion-reactor physics with high-power electron cyclotron resonance heating on ASDEX Upgrade

The electron cyclotron resonance heating (ECRH) system of the ASDEX Upgrade tokomak has been upgraded over the last 15 years from a 2MW, 2 s, 140 GHz system to an 8MW, 10 s, dual frequency system (105/140 GHz). The power exceeds the L/H power threshold by at least a factor of two, even for high densities, and roughly equals the installed ion cyclotron range of frequencies power. The power of both wave heating systems together (>10MW in the plasma) is about half of the available neutral beam injection (NBI) power, allowing significant variations of torque input, of the shape of the heating profile and of Qe/Qi, even at high heating power. For applications at a low magnetic field an X3-heating scheme is routinely in use. Such a scenario is now also forseen for ITER to study the first H-modes at one third of the full field. This versatile system allows one to address important issues fundamental to a fusion reactor: H-mode operation with dominant electron heating, accessing low collisionalities in full metal devices (also related to suppression of edge localized modes with resonant magnetic perturbations), influence of Te/Ti and rotational shear on transport, and dependence of impurity accumulation on heating profiles. Experiments on all these subjects have been carried out over the last few years and will be presented in this contribution. The adjustable localized current drive capability of ECRH allows dedicated variations of the shape of the q-profile and the study of their influence on non-inductive tokamak operation (so far at q$_{95}$>5.3). The ultimate goal of these experiments is to use the experimental findings to refine theoretical models such that they allow a reliable design of operational schemes for reactor size devices. In this respect, recent studies comparing a quasi-linear approach (TGLF) with fully non-linear modeling (GENE) of non-inductive high-beta plasmas will be reported