Monte Carlo simulation of ICRF discharge initiation in ITER

Discharges produced and sustained by ion cyclotron range of frequency (ICRF) waves in absence of plasma current will be used on ITER for (ion cyclotron-) wall conditioning (ICWC). The here presented simulations aim at ensuring that the ITER ICRH& CD system can be safely employed for ICWC and at finding optimal parameters to initiate the plasma. The 1D Monte Carlo code RFdinity1D3V was developed to simulate ICRF discharge initiation. The code traces the electron motion along one toroidal magnetic field line, accelerated by the RF field in front of the ICRF antenna. Electron collisions in the calculations are handled by a Monte Carlo procedure taking into account their energies and the related electron collision cross sections for collisions with H-2, H-2(+) and H+. The code also includes Coulomb collisions between electrons and ions (e - e; e - H-2(+); e - H+). We study the electron multiplication rate as a function of the RF discharge parameters (i) antenna input power (0.1-5MW), and (ii) the neutral pressure (H-2) for two antenna phasing (monopole [0000]-phasing and small dipole [0 pi 0 pi]-phasing). Furthermore, we investigate the electron multiplication rate dependency on the distance from the antenna straps. This radial dependency results from the decreasing electric amplitude and field smoothening with increasing distance from the antenna straps. The numerical plasma breakdown definition used in the code corresponds to the moment when a critical electron density nec for the low hybrid resonance (omega = omega(LHR)) is reached. This numerical definition was previously found in qualitative agreement with experimental breakdown times obtained from the literature and from experiments on the ASDEX Upgrade and TEXTOR

Modelling of the ICRF induced E x B convection in the scrape-off-layer of ASDEX Upgrade

In magnetic controlled fusion devices, plasma heating with radio-frequency (RF) waves in the ion cyclotron (IC) range of frequency relies on the electric field of the fast wave to heat the plasma. However, the slow wave can be generated parasitically. The electric field of the slow wave can induce large biased plasma potential (DC potential) through sheath rectification. The rapid variation of the rectified potential across the equilibrium magnetic field can cause significant convective transport (E x B drifts) in the scrape-off layer (SOL). In order to understand this phenomenon and reproduce the experiments, 3D realistic simulations are carried out with the 3D edge plasma fluid and kinetic neutral code EMC3-Eirene in ASDEX Upgrade. For this, we have added the prescribed drift terms to the EMC3 equations and verified the 3D code results against the analytical ones in cylindrical geometry. The edge plasma potential derived from the experiments is used to calculate the drift velocities, which are then treated as input fields in the code to obtain the final density distributions. Our simulation results are in good agreement with the experiments

Progress from ASDEX Upgrade experiments in preparing the physics basis of ITER operation and DEMO scenario development

An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L-H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate

On the mechanisms governing gas penetration into a tokamak plasma during a massive gas injection

A new 1D radial fluid code, IMAGINE, is used to simulate the penetration of gas into a tokamak plasma during a massive gas injection (MGI). The main result is that the gas is in general strongly braked as it reaches the plasma, due to mechanisms related to charge exchange and (to a smaller extent) recombination. As a result, only a fraction of the gas penetrates into the plasma. Also, a shock wave is created in the gas which propagates away from the plasma, braking and compressing the incoming gas. Simulation results are quantitatively consistent, at least in terms of orders of magnitude, with experimental data for a D 2 MGI into a JET Ohmic plasma. Simulations of MGI into the background plasma surrounding a runaway electron beam show that if the background electron density is too high, the gas may not penetrate, suggesting a possible explanation for the recent results of Reux et al in JET (2015 Nucl. Fusion 55 093013)

Impact of ICRF on the scrape-off layer and on plasma wall interactions: From present experiments to fusion reactor

Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma Wall Interactions) in ASDEX Upgrade (AUG), Alcator C-Mod, and JET-ILW are reviewed. Capabilities to diagnose and model the effect of DC biasing and associated impurity production at active antennas and on magnetic field connections to antennas are described. The experiments show that ICRF near-fields can lead not only to E×B convection, but also to modifications of the SOL density, which for Alcator C-Mod are limited to a narrow region near antenna. On the other hand, the SOL density distribution along with impurity sources can be tailored using local gas injection in AUG and JET-ILW with a positive effect on reduction of impurity sources. The technique of RF image current cancellation at antenna limiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna. Multiple observations confirmed the reduction of the impact of ICRF on the SOL and on total impurity production when the ratio of the power of the central straps to the total antenna power is in the range 0.6<P$_{cen}$/P$_{total}$<0.8. Near-field calculations indicate that this fairly robust technique can be applied to the ITER ICRF antenna, enabling the mode of operation with reduced PWI. On the contrary, for the A2 antenna in JET-ILW the technique is hindered by RF sheaths excited at the antenna septum. Thus, in order to reduce the effect of ICRF power on PWI in a future fusion reactor, the antenna design has to be optimized along with design of plasmafacing components