Design and first applications of the ITER integrated modelling & analysis suite

International audienceThe ITER Integrated Modelling & Analysis Suite (IMAS) will support both plasma operationand research activities on the ITER tokamak experiment. The IMAS will be accessible toall ITER members as a key tool for the scientific exploitation of ITER. The backbone of theIMAS infrastructure is a standardized, machine-generic data model that represents simulatedand experimental data with identical structures. The other outcomes of the IMAS design andprototyping phase are a set of tools to access data and design integrated modelling workflows,as well as first plasma simulators workflows and components implemented with variousdegrees of modularity

METIS: a fast integrated tokamak modelling tool for scenario design

International audienceMETIS is a numerical code aiming at fast full tokamak plasma analyses and predictions. It combines 0-D scaling-law normalised heat and particle transport with 1-D current diffusion modelling and 2-D equilibria. It contains several heat, particle and impurities transport models, as well as heat, particle, current and momentum sources, which allow faster than real time scenario simulations. This paper gives a first comprehensive description of the METIS suite: overall structure of the code, main available models, details on the simulation workflow and numerical implementation. Some examples of applications to the analysis of experimental discharges and the predictions of ITER scenarios are also given

Development of RF heating and current drive systems for long pulse operation on Tore Supra

International audienc

Contribution of Tore Supra in preparation of ITER

Tore Supra routinely addresses the physics and technology of very long-duration plasma discharges, thus bringing precious information on critical issues of long pulse operation of ITER. A new ITER relevant lower hybrid current drive (LHCD) launcher has allowed coupling to the plasma a power level of 2.7 MW for 78 s, corresponding to a power density close to the design value foreseen for an ITER LHCD system. In accordance with the expectations, long distance (10 cm) power coupling has been obtained. Successive stationary states of the plasma current profile have been controlled in real-time featuring (i) control of sawteeth with varying plasma parameters, (ii) obtaining and sustaining a 'hot core' plasma regime, (iii) recovery from a voluntarily triggered deleterious magnetohydrodynamic regime. The scrape-off layer (SOL) parameters and power deposition have been documented during L-mode ramp-up phase, a crucial point for ITER before the X-point formation. Disruption mitigation studies have been conducted with massive gas injection, evidencing the difference between He and Ar and the possible role of the q = 2 surface in limiting the gas penetration. ICRF assisted wall conditioning in the presence of magnetic field has been investigated, culminating in the demonstration that this conditioning scheme allows one to recover normal operation after disruptions. The effect of the magnetic field ripple on the intrinsic plasma rotation has been studied, showing the competition between turbulent transport processes and ripple toroidal friction. During dedicated dimensionless experiments, the effect of varying the collisionality on turbulence wavenumber spectra has been documented, giving new insight into the turbulence mechanism. Turbulence measurements have also allowed quantitatively comparing experimental results with predictions by 5D gyrokinetic codes: numerical results simultaneously match the magnitude of effective heat diffusivity, rms values of density fluctuations and wavenumber spectra. A clear correlation between electron temperature gradient and impurity transport in the very core of the plasma has been observed, strongly suggesting the existence of a threshold above which transport is dominated by turbulent electron modes. Dynamics of edge turbulent fluctuations has been studied by correlating data from fast imaging cameras and Langmuir probes, yielding a coherent picture of transport processes involved in the SOL. Corrections were made to this article on 6 January 2012. Some of the letters in the text were missing

Lower Hybrid Current Drive in Tore Supra and Jet

Recent Lower Hybrid Current Drive (LHCD) experiments in TORE SUPRA and JET are reported. Large multijunction launchers have allowed the coupling of 5 MW to the plasma for several seconds with a maximum of 3.8 kw/cm2. Measurements of the scattering matrices of the antennae show good agreement with theory. The current drive efficiency in TORE SUPRA is about 0.2 x 10(20) Am-2/W with LH power alone and reaches 0.4 x 10(20) Am-2/W in JET thanks to a high volume-averaged electron temperature (1.9 keV) and also to a synergy between Lower Hybrid and Fast Magnetosonic Waves. At N(e)BAR = 1.5 x 10(19) m-3 in TORE SUPRA, sawteeth are suppressed and m = 1 MHD oscillations the frequency of which clearly depends on the amount of LH power are observed on soft x-rays, and also on non-thermal ECE. In JET ICRH produced sawtooth-free periods are extended by the application of LHCD (2.9 s. with 4 MW ICRH) and current profile broadening has been clearly observed consistent with off-axis fast electron populations. LH power modulation experiments performed in TORE SUPRA at N(e)BAR = 4 x 10(19) m-3 show a delayed central electron heating despite the off-axis creation of suprathermal electrons, thus ruling out the possibility of a direct heating through central wave absorption. A possible explanation in terms of anomlous fast electron transport and classical slowing down would yield a diffusion coefficient of the order of 10 m2/s for the fast electrons. Other interpretations such as an anomalous heat pinch or a central confinement enhancement cannot be excluded. Finally, successful pellet fuelling of a partially LH driven plasma was obtained in TORE SUPRA, 28 successive pellets allowing the density to rise to N(e)BAR = 4 x 10(19) m-3. This could be achieved by switching the LH power off for 90 ms before each pellet injection, i.e. without modifying significantly the current density profile

Science and technology research and development in support to ITER and the Broader Approach at CEA

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented