[en] In thermonuclear fusion devices, controlling the Deuterium-Tritium fuel density and exhausting the Helium ashes is a crucial point. This is achieved by fuelling the discharges by different methods (gas puffing and pellet injection are the most commonly used) and by implementing pumping devices at the plasma periphery. These two issues are treated in this work, both from an experimental and a modelling point of view, using the neutral transport code EIRENE as main tool for our studies. As far as pumping is concerned, we have modelled the outboard pump limiter of the Tore Supra tokamak with the EIRENE code to which we coupled a plasma module specially developed to simulate the neutrals and the plasma in a coherent way. This allowed to validate the code against experimental data. As far as plasma fuelling is concerned, we present here an original method: the supersonic pulsed gas injection (SPGI). This intermediate method between conventional gas puff (GP) and pellet injection was designed and tested at Tore Supra. It consists of injecting very dense and short gas puffs at high speed into the plasma. Experimentally, SPGI was found to have a better fuelling efficiency than GP and to lead to a strong plasma cooling. The mechanisms responsible for this improved efficiency are analysed by modelling, using the EIRENE code to determine the ionisation source and a 1 D transport model to reproduce the plasma density response. At last, an extrapolation of the present injector is presented, discussing the possibility to obtain a radial drift of the injected matter as observed in the case of high field side pellet injection. (author)

[en] Full text: The COMPASS tokamak is one of the present devices operating with an ITER-like plasma shape. Its flexibility combined to an extensive set of diagnostics and NBI heating allow to address a broad range of key areas in support of the worldwide fusion programme such as H-mode, MHD, RAE, disruptions, PWI. The recent results obtained in COMPASS addressing these key issues are reviewed here. The control and characterization of the L-H transition and the pedestal physics represent a large part of the COMPASS scientific programme. Recycling and actuators such as the X-point height play a significant role in accessing H-mode. GAMs oscillating at frequencies 25–40 kHz are observed in L-mode discharges, increasing with the ion mass and with a decreasing amplitude from D to H plasmas. COMPASS also contributes to multimachine databases with pedestal and SOL width scalings studies. Using perturbation coils, the influence of 3D fields on the strike-points splitting, ELM control and transport is reported. The MHD modes studies mainly concern the plasma interaction with RMPs, the characterization of Alfvén-like modes and disruption/mitigation experiments. High frequency quasi-coherent oscillations (ranging from 200 kHz to above 1 MHz) that follow Alfvénic frequency scaling are observed in ohmic discharges. An extensive experimental study of MHD effects in losses of runaway electrons has been performed. In the field of disruptions, an intermachine empirical scaling of critical magnetic disruption precursors has been developed, as well as the study of the disruptions toroidal asymmetry. The exhaust and plasma-material interaction studies in COMPASS contributed to the ITER divertor monoblocks design as part of the ITPA. Power deposition on leading edge was investigated both experimentally (inner-wall limiters with gaps and leading edges viewed by a high-resolution IR camera) and numerically (PIC simulations), with the latter reproducing well this experiment and the recent lamella melting experiment on JET. The ITER monoblocks shaping was also investigated in the frame of an ITER contract. Comparison with the deposited power from ion orbit calculations are consistent and confirm results presented at the previous IAEA FEC. However, the role of the E-field and the contribution from the electrons on the total power flux accounted in PIC calculations predict marginally lower power. (author)

[en] Plasma-facing divertors and limiters are armoured with castellated tiles to withstand intense heat fluxes. Recent experimental studies show that a non-negligible amount of deuterium is deposited in the gaps between tiles. We present here a numerical study of plasma deposition in this critical region. For this purpose we have developed a particle-in-cell code with realistic boundary conditions determined from kinetic calculations. We find a strong asymmetry of plasma deposition into the gaps. A significant fraction of the plasma influx is expelled from the gap to be deposited on the leading edge of the downstream tile

[en] A fueling system by supersonic pulsed gas injection has been installed on the high field side of Tore Supra. First results are encouraging, demonstrating a fueling efficiency four times higher than that of conventional gas puff. One-dimensional modeling shows that the increase of efficiency is linked to the short injection time and to the prompt cooling of the plasma edge consecutive to the massive injection of matter. Improvements of the system could lead to the formation of a high-β blob which could experience a drift down the magnetic field, analogously to pellet injection, thus further increasing the fueling efficiency of the method

[en] The tunnel probe is a concave Langmuir probe designed to operate in strongly magnetized plasma. Due to its shape, the tunnel probe is immune to sheath expansion effects and thus provides absolutely calibrated measurements of the parallel ion current density. A two-dimensional, self-consistent kinetic model is employed to model the flow of charges within the cavity of the tunnel probe. The calculation predicts that the distribution of the ion flux onto the inner conductors depends on the electric field inside the tunnel, which in turn depends on the electron temperature. Therefore, if the tunnel is divided into two negatively biased collectors, it is possible to use the simulation results to determine the electron temperature from the measured ion current ratio. This means that a DC-biased tunnel probe can be used to provide fast, simultaneous measurements of the parallel ion current density and the electron temperature without collecting a single electron. Measurements in the CASTOR and Tore Supra tokamaks agree well with the numerical simulations. (paper)

No abstract available

[en] The ratio of electron and ion saturation currents measured by probes embedded in the plasma-facing components (PFCs) in magnetised plasmas is typically lower than predicted by Langmuir theory. In the past, several works targeted various effects influencing the ion saturation current, which can be larger than expected, especially if the magnetic field is at a grazing angle with respect to the probe surface. In this contribution, we focus on the electron current, which can be reduced when the magnetic field line intercepting the probe passes through the magnetic pre-sheath of the nearby PFC. In such a case an effective potential well forms and repels a fraction of the electrons incoming from the plasma. Experimental results obtained by a tiltable limiter at the DITE tokamak are revisited and reproduced by means of 2D3V particle-in-cell simulations. The reduction of the electron current is indeed observed both in simulations and in experiment. This mechanism may also explain why some divertor biasing experiments did not produce flows in the divertor region. (paper)

[en] The effect of the use of different electrode materials for edge-plasma biasing on plasma density and floating-potential profile modifications has been recently investigated on the CASTOR tokamak. Dependences of biasing current, edge plasma density and floating potential on biasing voltage have been measured. Induced relaxation events and fast (10-50 kHz) oscillations are shown and surface changes of the biasing electrodes are discussed in the paper. (author)

[en] The evaluation of probe measurements in the swept regime can be performed via a fit to the ion branch of the current-voltage (I–V) characteristic, providing the most relevant plasma parameters such as the ion saturation current, the electron temperature and the floating potential. In this paper, we present a parametric study based upon the existing flush-mounted probe setup briefly operated at the COMPASS tokamak, aiming to compare relevant I–V formulas widely used for such analysis. Selected four-parametric fit descriptions were applied to synthetic data obtained by simulation of the probe in a particle-in-cell model, SPICE, both in simplified 2D and more complex 3D geometries. Plasma parameters recovered by the fit were compared to the input from the model and the precision of this recovery was mapped to a wide range of parameters, especially with respect to the ratio of the probe size d _pin to the Larmor radius r _L. Results show that while the electron temperature and the floating potential can be obtained quite precisely regardless of the method, the ion saturation current can be identified incorrectly, mainly due to the fact that the sheath expansion description by optical approximation is not sufficient for almost grazing field line angles of incidence, especially when d _pin ∼ r _L. We show, however, that operation of flush-mounted probes can be fruitful especially in high-field devices. Additionally, we have explored the sensitivity on the available bias potential range, showing that for proper analysis of I–V characteristic using four-parametric fit, the lower bound of the fit range should be at least $4k_{\mathrm{B}}{T}_{\mathrm{e}}/e$ below the floating potential. If this condition cannot be satisfied, three-parametric fit in the range close to the floating potential can still produce results with acceptable precision. (paper)

[en] The COMPASS tokamak (R = 0.56 m, a = 0.2 m, B_T = 1.3 T, I_p ~ 300 kA, pulse duration 0.4 s) operates in ITER-like plasma shape in H-mode with Type-I ELMs. In 2019, we plan to install into the divertor a test target based on capillary porous system filled with liquid lithium/tin. This single target will be inclined toroidally in order to be exposed to ITER-relevant surface heat flux (20 MW/m²). Based on precisely measured actual heat fluxes, our simulations predict (for 45° inclination, without accounting for the lithium vapor shielding) the surface temperature rises up to 700°C within 120 ms of the standard ELMy H-mode heat flux with ELM filaments reaching hundreds MW/m². Significant lithium vaporization is expected. The target surface will be observed by spectroscopy, fast visible and infrared cameras. The scientific program will be focused on operational issues (redeposition of the evaporated metal, ejection of droplets, if any) as well as on the effect on the plasma physics (improvement of plasma confinement, L–H power threshold, Z_eff, etc.). After 2024, a closed liquid divertor may be installed into the planned COMPASS Upgrade tokamak (R = 0.84 m, a = 0.3 m, B_T = 5 T, I_p = 2 MA, P_in = 8 MW, pulse duration ~2 s) with ITER-relevant heat fluxes loading the entire toroidal divertor.