1 OV/P-3 Overview of Recent COMPASS Activities R. Dejarnac1 and the COMPASS team1,* 1Institute of Plasma Physics, Czech Academy of Sciences, Prague, Czech Republic * see appendix E-mail contact of main author: [email protected] Abstract. The COMPASS tokamak is one of the present devices operating with an ITER-like plasma shape. Its flexibility due to its small size combined to an extensive set of edge diagnostics and NBI heating allow to address a broad range of key areas in support of the worldwide fusion programme such as H-mode physics, MHD, runaways electrons, disruption studies, plasma-wall interactions. The recent results obtained in COMPASS addressing these key issues are reviewed here. 1. Introduction 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 neutral beam injection heating allow to address a broad range of key areas in support of the worldwide fusion programme such as H-mode, MHD, runaway electrons, disruptions or plasma-surface interaction studies. 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. A geodesic acoustic mode oscillating at frequencies 25-40 kHz is observed in L-mode discharges, strongly varying with NBI heating in co- or counter-current injections. COMPASS also contributes to multi- machine databases with pedestal and scrape-off layer (SOL) width scaling 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 Resonant Magnetic Perturbations (RMPs), the characterization of Alfven-like modes and disruption/mitigation experiments. High frequency quasi-coherent oscillations (ranging from 200 kHz to above 1 MHz) that follow Alfvenic 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 inter-machine 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 International Tokamak Physics Activity (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 (particle-in-cell simulations), with the latter reproducing well this experiment. 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 by J. P. Gunn et al.. 2 OV/P-3 2. The COMPASS tokamak Since its relocation in the Institute of Plasma Physics of the Czech Academy of Sciences (IPP CAS) in 2006 [1], COMPASS is equipped with new power supply and control systems, two neutral beams injectors (NBI) and an extensive set of diagnostics. All this together, combined with an ITER-like plasma shape, enables to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scaling. 2.1. Machine description and operation scenarios The COMPASS tokamak is a compact experimental device (R = 0.56 m, a = 0.2 m) operating in a divertor plasma configuration with ITER-like plasma cross-section (1 : 10 in linear dimension) [1,2]. The plasma current can go up to 400 kA, the toroidal magnetic field in the range of 0.8–2.1 T, elongation up to 1.8 and different working gases can be used: H, D or He. COMPASS can generate various plasma configurations: circular, elongated or single null X- point (SND) for both toroidal directions of BT and Ip. The vacuum vessel (VV) of COMPASS is equipped with an open divertor covered by carbon tiles. Two Neutral Beam Injectors operating with deuterium/hydrogen beams (Eb = 40 keV, port-through power PNBI < 0.35 MW each, pulse length < 0.3 s) are used for additional plasma heating. The beams are injected tangentially (on-axis, co-injection at present). Extensive wall conditioning methods (baking at 150 °C, boronization by carborane vapors in He glow discharge) are used to minimize the deuterium inventory and impurity influx. Upon its re-installation in IPP Prague, both ohmic and NBI assisted H-modes have been successfully generated after the above mentioned conditioning. In ohmic regime, the H-mode LH is achieved when the plasma current exceeds approximately IP ~ 220 kA (at BT = 1.2 T). The L–H power threshold PLH has a minimum at line-average density in the range of (3.5– 4) × 1019 m−3. At these densities the beams of neutral atoms generated by NBIs are already fully absorbed with a negligible shine-through, however, NBI allows to heat plasma also at higher densities. The L–H transition is usually followed either by small Type-III ELMs with frequencies in the range of 400–1500 Hz or ELM-free period. At high plasma currents or in case of NBI-heated plasma, large Type-I ELMs with typical frequency in the approximate range of 80–400 Hz are generated. 2.2. Diagnostics The COMPASS tokamak is equipped with an extensive set of diagnostics with high temporal and spatial resolution to focus mainly on edge and SOL plasma studies. The core and edge plasma regions are diagnosed using a High Resolution Thomson Scattering (HRTS) system [2,8], based on two Nd:YAG lasers (1.5 J each, 30 Hz) in conjunction with two detection systems, to provides ne and Te profiles with a temporal resolution of 60 Hz and a spatial resolution of 3–4 mm in the vertical direction, corresponding to 2–3 mm (∼a/100) when mapped to the outer mid-plane (OMP). Fast edge density profiles are measured at OMP using Beam Emission Spectroscopy on Lithium beam operating at 60 keV with 4 μs chopping and spatial resolution ∼1 cm and also by using a fast reflectometer in K and Ka bands with ∆t∼ 30 μs. Six arrays of AXUV-based fast bolometers (10 eV < Eph < 10 keV) and two arrays of SXR detectors (1 keV < Eph < 20 keV) provide the tomographic reconstruction of the plasma emissivity with temporal resolution ∼1 μs. A set of 400 magnetic diagnostic coils can provide a full poloidal coverage at several toroidal locations. Local poloidal magnetic fields and their fluctuations are measured by a set 3 OV/P-3 of 16 internal partial Rogowski coils located at one toroidal position and three poloidal arrays of 24 Mirnov coils located at three different toroidal positions. These coils enable poloidally localized measurement of toroidal mode number n in the range of 1–4. Moreover, the tokamak chamber is fully covered by 104 saddle loops that also enable poloidally localized measurements of n in the range of 0–2 (on the low-field side in the range of 0–4). Magnetic data also provide input for the EFIT equilibrium reconstruction. Visible light is monitored by two fast EDICAM cameras with 450 fps in full-frame (1280 x 1024) up to 116 000 fps in reduced (16 × 16) resolutions. The first camera provides a general view of the plasma configuration (position, shape, wall interaction, etc.) via a tangential view. The second camera is used for monitoring of the divertor region. Additionally, two ultra-fast Photron cameras of APX-RS and SA5 with 3000/7000 fps in full (1024 × 1024) up to 250 kfps (128 × 16) / 1 Mfps (64 × 16) in reduced resolutions are used for observation of fast MHD events (ELM filamentary structure, Beta-induced Alfven Eingenmodes, Geodesic-Acoustic Modes, etc.). SOL physics is studied using two reciprocating manipulators (one horizontal at OMP and one vertical from the top of the vessel). The horizontal reciprocating manipulator (HRCP) is equipped by an interface compatible with the manipulator at ASDEX Upgrade and allows installation of wide variety of probe heads, such as Retarding Field Analyzer, E×B analyzer or several Langmuir probes (LP) and ball-pen probes (BPP). The Langmuir tips usually measure the floating potential Vfl and the ion saturation current Isat, whilst the BPPs, operating in floating mode, provide fast measurements of plasma potential φpl [3]. It was shown [3] that the difference between the signal of the BPP and the signal of an adjacent floating Langmuir tip is proportional to electron temperature VBPP − VLP = αTe. The coefficient α ≈ 2.2 was found by benchmarking the probe data by HRTS measurements. The principal advantage of Te measurement by combination of the BPP and floating LP is that the temporal resolution is limited mainly by the sampling frequency of the data acquisition system (5 MHz on COMPASS). Combining density and temperature measurements provide fast measurement of the power flux. The plasma in the divertor region is analyzed by a poloidal array of 39 LPs with a spatial resolution of approximately 5 mm and by an recently installed array of 55 BPPs and 110 LPs, located 135o toroidally away. 3. Edge and SOL physics 3.1. Effect of the X-point height on the L-H transition It is known that the underlying physics of the access to H-mode is still not fully understood. An empirical scaling law [4] based on global parameters such as density, toroidal field and size of the plasma is the only prediction for ITER to estimate the L- to H-mode power threshold.