M. Tsalas

FOM Institute AMOLF, Amsterdamo, North Holland, Netherlands

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Publications (49)67.07 Total impact

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    ABSTRACT: A reciprocating Langmuir probe was used to directly measure the behavior of turbulence and flows in the X-point region during transitions between low-(L) and high-confinement (H) mode in ASDEX Upgrade. The probe traverses the divertor horizontally in 140 ms, typically 2-5 cm below the X-point. Toroidal Mach number, density, floating potential (ϕf), and electron temperature (Te) are measured. In the regime accessible to the probe (Pinj<1.5 MW, line-integrated core density <4×1019 m-2), the L-H transition features an intermediate phase (I-phase), characterized by limit-cycle oscillations at 0.5-3 kHz [Conway et al., Phys. Rev. Lett. 106, 065001 (2011)]. The probe measurements reveal that this pulsing affects both the density and the toroidal Mach number. It is present in both the low-(LFS) and high-field sides (HFS) of the scrape-off layer, while high-amplitude broadband turbulence usually dominates the private-flux region. Profile comparisons between L-mode and I-phase show lower density in pulsing regions and small shifts in Te, directed oppositely on LFS and HFS, which are compensated by shifts in ϕf to yield a surprisingly unchanged plasma potential profile. Directly observed L-I-phase transitions reveal that the onset of the pulsing is preceded by a fast 50% density drop in the HFS X-point region. Back transitions to L-mode occur essentially symmetrically, with the pulsing stopping first, followed by a fast recovery to L-mode density levels in the divertor.
    03/2014; 21(4).
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    ABSTRACT: The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m−2. Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix.
    Nuclear Fusion 09/2013; 55:104003. · 2.73 Impact Factor
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    ABSTRACT: This paper reports on the recent assessment of the Ion Cyclotron Wall Conditioning (ICWC) technique for isotopic ratio control, fuel removal and recovery after disruptions, which has been performed on TORE SUPRA, TEXTOR, ASDEX Upgrade and JET. ICWC discharges were produced using the standard ICRF heating antennas of each device, at different frequencies and toroidal fields, either in continuous or pulsed mode. Intrinsic ICWC discharge inhomogeneities could be partly compensated by applying a small vertical magnetic field, resulting in the vertical extension of the discharge in JET and TEXTOR. The conditioning efficiency was assessed from the flux of desorbed and retained species, measured by means of mass spectrometry. In Helium ICWC discharges, fuel removal rates between 1016D.m-2.s-1 to 3.1017D.m-2.s-1 were measured, with a linear dependence on the coupled RF power and on the He + density. ICWC scenarios have been developed in D or H plasmas for isotopic exchange. The H (or D) outgassing was found to increase with the D (resp. H) partial pressure. In continuous mode, wall retention is on the average two to ten times higher than desorption , due to the high reionization probability of desorbed species in ICWC discharges, where the electron density is about 1018m-3. Retention can be minimized in pulsed ICWC discharges without severely reducing outpumping. Pulsed He-ICWC discharges have been successfully used on TORE SUPRA to recover normal operation after disruptions, when subsequent plasma initiation would not have been possible without conditioning.
    Journal of Nuclear Materials 08/2013; 415:S1021–S1028. · 2.02 Impact Factor
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    ABSTRACT: Detailed experimental studies of ion heat transport are carried out in JET to explore the Te/Ti dependence of ion heat transport in ITER relevant range of parameters (Te/Ti ≥ 1) using low rotation plasmas with dominant ion cyclotron resonance heating to avoid the coupling of the effects of Te/Ti and rotation which affected previous experiments. This experimental setup has led to an accurate determination of the ion temperature gradient (ITG) threshold at varying Te/Ti, offering unique opportunities for validation of the well-established theory of ITG driven modes. A rather mild decrease in threshold with increasing Te/Ti in the interval of ITER interest was found. The new experimental result has found good agreement with theoretical predictions based on quasi-linear fluid and linear gyrokinetic models.
    Plasma Physics and Controlled Fusion 04/2013; 55(5):055003. · 2.37 Impact Factor
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    ABSTRACT: Starting from recent JET experimental results that show a significant reduction of ion stiffness in the plasma core region due to plasma rotation in the presence of low magnetic shear, an experiment was carried out at JET in order to separate the role of rotation and rotation gradient in mitigating the ion stiffness level. Enhanced toroidal field ripple (up to 1.5%) and external resonant magnetic fields are the two mechanisms used to try and decouple the rotation value from its gradient. In addition, shots with reversed toroidal field and plasma current, yielding counter-current neutral beam injection, were compared with standard co-injection cases. These tools also allowed varying the rotation independently of the injected power. Shots with high rotation gradient are found to maintain their low stiffness level even when the absolute value of the rotation was significantly reduced. Conversely, high but flat rotation yields much less peaked ion temperature profiles than a peaked rotation profile with lower values. This behaviour suggests the rotation gradient as the main player in reducing the ion stiffness level. In addition, it is found that inverting the rotation gradient sign does not suppress its effect on ion stiffness.
    Plasma Physics and Controlled Fusion 01/2013; 55(2):025010. · 2.37 Impact Factor
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    ABSTRACT: The new full-metal ITER-like wall (ILW) at JET was found to have a profound impact on the physics of disruptions. The main difference is a significantly lower fraction (by up to a factor of 5) of energy radiated during the disruption process, yielding higher plasma temperatures after the thermal quench and thus longer current quench times. Thus, a larger fraction of the total energy was conducted to the wall resulting in larger heat loads. Active mitigation by means of massive gas injection became a necessity to avoid beryllium melting already at moderate levels of thermal and magnetic energy (i.e. already at plasma currents of 2 MA). A slower current quench, however, reduced the risk of runaway generation. Another beneficial effect of the ILW is that disruptions have a negligible impact on the formation and performance of the subsequent discharge.
    Plasma Physics and Controlled Fusion 11/2012; 54(12):124032. · 2.37 Impact Factor
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    ABSTRACT: The X-point reciprocating Langmuir probe in ASDEX Upgrade has been recently upgraded with a new drive system and higher-bandwidth measurement electronics. The horizontal plunge direction allows the probe to penetrate through both LFS and HFS at the level of the X-point, thus completely covering the divertor entrance. We present first measurements of density, temperature and flow profiles in different L- and H-mode plasmas close to the transition. The flows are generally larger on the HFS, where the plasma is also denser and colder. Significant differences in temperature and flow patterns are observed in different plasma conditions. Ideas for upgrading the probe shaft to perform LFS and HFS measurements simultaneously for the characterization of transients will be presented.
    10/2012;
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    Plasma Physics and Controlled Fusion 09/2012; · 2.37 Impact Factor
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    ABSTRACT: This paper summarizes the operational experience of the ion cyclotron resonant frequency (ICRF) ITER-like antenna on JET aiming at substantially increasing the power density in the range of the requirements for ITER combined with load resiliency. An in-depth description of its commissioning, operational aspects and achieved performances is presented.
    Plasma Physics and Controlled Fusion 06/2012; 54(7):074012. · 2.37 Impact Factor
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    ABSTRACT: A new Thomson scattering diagnostic is proposed for the study of fast plasma dynamics in the pedestal of ASDEX Upgrade. The diagnostic will measure electron temperature and density profiles over a ∼ 3 cm wide area in the edge transport barrier region, with ∼ 1–2 mm spatial resolution and ∼ 10 kHz sampling rate. A challenging goal of the project is the study of the bootstrap current in the plasma pedestal by measuring the distortion and shift of the electron distribution along the toroidal direction. Expected spatial and time resolutions of the current density measurements are ∼ 3 mm and ∼ 1 ms correspondingly. The new diagnostic will be used to study the fast dynamic behaviour of the pedestal bootstrap current, where models indicate that it plays a key role in regulating edge stability, e.g. during ELMs. The diagnostic design is based on the intra-cavity multi-pass system currently in operation in TEXTOR, which uses a probing ruby laser, a grating spectrometer and two fast CMOS cameras for scattered light detection, and has achieved measuring accuracies of the order of ∼ 1% for ne and ∼ 2% for Te. Parts of that system will be reused in ASDEX Upgrade (some with significant modifications), but the laser multi-pass and light collection systems are entirely redesigned. Restrictions in space and line-of-sight availability have led to the adoption of a design which uses in-vessel multi-pass mirrors and light collection optics, requiring a number of innovative technical solutions to permit remote laser alignment and light collection. We give an overview of the project, discuss the underlying physics basis and present a number of technical solutions employed.
    Journal of Instrumentation - J INSTRUM. 01/2012; 7(03).
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    ABSTRACT: Detailed experimental studies of ion heat transport have been carried out in JET exploiting the upgrade of active charge exchange spectroscopy and the availability of multi-frequency ion cyclotron resonance heating with 3He minority. The determination of ion temperature gradient (ITG) threshold and ion stiffness offers unique opportunities for validation of the well-established theory of ITG driven modes. Ion stiffness is observed to decrease strongly in the presence of toroidal rotation when the magnetic shear is sufficiently low. This effect is dominant with respect to the well-known ωE×B threshold up-shift and plays a major role in enhancing core confinement in hybrid regimes and ion internal transport barriers. The effects of Te/Ti and s/q on ion threshold are found rather weak in the domain explored. Quasi-linear fluid/gyro-fluid and linear/non-linear gyro-kinetic simulations have been carried out. Whilst threshold predictions show good match with experimental observations, some significant discrepancies are found on the stiffness behaviour.
    Plasma Physics and Controlled Fusion 11/2011; 53(12):124033. · 2.37 Impact Factor
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    ABSTRACT: Several parametric scans have been performed to study momentum transport on JET. A neutral beam injection modulation technique has been applied to separate the diffusive and convective momentum transport terms. The magnitude of the inward momentum pinch depends strongly on the inverse density gradient length, with an experimental scaling for the pinch number being -Rvpinch/χ = 1.2R/Ln + 1.4. There is no dependence of the pinch number on collisionality, whereas the pinch seems to depend weakly on q-profile, the pinch number decreasing with increasing q. The Prandtl number was not found to depend either on R/Ln, collisionality or on q. The gyro-kinetic simulations show qualitatively similar dependence of the pinch number on R/Ln, but the dependence is weaker in the simulations. Gyro-kinetic simulations do not find any clear parametric dependence in the Prandtl number, in agreement with experiments, but the experimental values are larger than the simulated ones, in particular in L-mode plasmas. The extrapolation of these results to ITER illustrates that at large enough R/Ln > 2 the pinch number becomes large enough (>3–4) to make the rotation profile peaked, provided that the edge rotation is non-zero. And this rotation peaking can be achieved with small or even with no core torque source. The absolute value of the core rotation is still very challenging to predict partly due to the lack of the present knowledge of the rotation at the plasma edge, partly due to insufficient understanding of 3D effects like braking and partly due to the uncertainties in the extrapolation of the present momentum transport results to a larger device.
    Nuclear Fusion 11/2011; 51(12):123002. · 2.73 Impact Factor
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    ABSTRACT: New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.
    Physical Review Letters 09/2011; · 7.73 Impact Factor
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    ABSTRACT: Modification of the sawtooth period through ion cyclotron resonance frequency (ICRF) heating and current drive has been demonstrated in a number of experiments. The effect has been seen to depend critically on the location of the ICRF absorption region with respect to the q = 1 flux surface. Consequently, for ICRF to be a viable tool for sawtooth control, one must be able to control the ICRF absorption location in real time so as to follow variations in the location of the q = 1 surface. To achieve this, the JET ICRF system has been modified to allow the JET real time central controller to control the frequency of the ICRF generators. An algorithm for real time determination of the sawtooth period has been developed and a closed loop controller, which modifies the frequency of the ICRF generators to bring the measured sawtooth period to the desired reference value, has been implemented. This paper shows the first experimental demonstration of closed loop sawtooth period control by real time variation of the ICRF wave frequency.
    Nuclear Fusion 06/2011; 51(7):073032. · 2.73 Impact Factor
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    ABSTRACT: Preparatory experiments for the ITER-Like Wall in JET were carried out to simulate the massive Be first wall by a thin Be layer, induced by evaporation of about 2.0 g Be, and to study its impact on fuel retention and divertor radiation with reduced C content and N seeding. Residual gas analysis reveals a reduction of hydrocarbons by one order of magnitude and of O by a factor of 5 in the partial pressure owing to the evaporation. The evolution of wall conditions, impurity fluxes and divertor radiation have been studied in ELMy H-mode plasmas (Bt = 2.7 T, Ip = 2.5 MA, Paux = 16 MW) whereas a non-seeded reference discharge was executed prior to the evaporation.The in situ measured Be flux at the midplane increased by about a factor of 40 whereas the C flux decreased by ~50% in the limiter phase of the first discharge with respect to the reference, but erosion of the Be layer and partial coverage with C takes place quickly. To make best use of the protective Be layer, only the first four discharges were employed for a gas balance analysis providing a D retention rate of 1.94 × 1021 D s−1 which is comparable to rates with C walls. But the Be evaporation provides a non-saturated surface with respect to D and short term retention is not negligible in the balance; the measured retention is overestimated with respect to steady-state conditions like that of the ILW. Moreover, C was only moderately reduced and co-deposition of fuel with eroded Be and C occurs.The lower C content leads to a minor reduction in divertor radiation as the reference phase prior to seeding indicates. N adds to the radiation of D and the remaining C, and the N content rises due to the legacy effect which has been quantified by gas balance to be 30% of the injected N. C radiation increases with exposure time, and both contributors cause an increase in the radiated fraction in the divertor from 50% to 70%. The radiation pattern suggests that N dominates the increase in the first discharges though C is still the dominating radiator. Therefore, the validity of a proxy of the Be first wall by a thin Be layer is limited and restricted to plasma operation directly after the Be evaporation.
    Nuclear Fusion 05/2011; 51(7):073007. · 2.73 Impact Factor
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    ABSTRACT: Accurate and validated tools for calculating toroidal momentum sources are necessary to make reliable predictions of toroidal rotation for current and future experiments. In this work we present the first experimental validation of torque profile calculation from neutral beam injection (NBI) under toroidal field ripple. We use discharges from a dedicated experimental session on JET where neutral beam modulation technique is used together with time-dependent torque calculations from ASCOT code for making the benchmark. Good agreement between simulations and experimental results is found.
    Plasma Physics and Controlled Fusion 05/2011; 53(8):085005. · 2.37 Impact Factor
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    ABSTRACT: Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0–2.0 × 1021 D s−1 were obtained as references in accompanied gas balance studies.
    Journal of Nuclear Materials 01/2011; · 2.02 Impact Factor
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    ABSTRACT: A new pellet injection system was installed at JET designed for both fuelling and ELM pacing. The purpose of the pacing section was to validate pellet ELM pacing as a suitable tool for ELM mitigation in ITER. Pellet pacing was confirmed at the large size scale of JET. The dynamics of triggered ELMs was investigated with respect to their spontaneous counterparts. Triggered ELMs show features also typical for spontaneous ELMs in several operational regimes. Since none of these regimes was unsettled by the pellets this is a strong hint for compatibility with other plasma control tools. Observations and modelling results indicate the ELM triggering occurs by the evolution of the pellet ablation plasmoid into the first ELM filament followed by a poloidal spread of the instability. An ELM obviously can be forced by a pellet due to the strong local perturbation imposed already under unusual onset conditions but then evolves like any ELM typical for the corresponding plasma regime. For tool optimization the pellet mass and hence the convective confinement losses imposed have to be minimized. In our experiments, a lower mass threshold was observed for the first time. It has been found that to reliably trigger an ELM the pellet needs to be sufficiently large (and fast) to penetrate close to the pedestal top. Recent investigations are clear steps forward to validate the pellet pacing approach for ITER.
    Nuclear Fusion 01/2011; · 2.73 Impact Factor
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    ABSTRACT: The scan of ion cyclotron resonant heating (ICRH) power has been used to systematically study the pump out effect of central electron heating on impurities such as Ni and Mo in H-mode low collisionality discharges in JET. The transport parameters of Ni and Mo have been measured by introducing a transient perturbation on their densities via the laser blow off technique. Without ICRH Ni and Mo density profiles are typically peaked. The application of ICRH induces on Ni and Mo in the plasma centre (at normalized poloidal flux rho = 0.2) an outward drift approximately proportional to the amount of injected power. Above a threshold of ICRH power of about 3 MW in the specific case the radial flow of Ni and Mo changes from inwards to outwards and the impurity profiles, extrapolated to stationary conditions, become hollow. At mid-radius the impurity profiles become flat or only slightly hollow. In the plasma centre the variation of the convection-to-diffusivity ratio v/D of Ni is particularly well correlated with the change in the ion temperature gradient in qualitative agreement with the neoclassical theory. However, the experimental radial velocity is larger than the neoclassical one by up to one order of magnitude. Gyrokinetic simulations of the radial impurity fluxes induced by electrostatic turbulence do not foresee a flow reversal in the analysed discharges.
    Nuclear Fusion 01/2011; · 2.73 Impact Factor
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    ABSTRACT: Wall conditioning techniques applicable in the presence of permanent toroidal magnetic field will be required for the operation of ITER, in particular for recovery from disruptions, vent and air leak, isotopic ratio control, recycling control and mitigation of the tritium inventory build-up. Ion Cyclotron Wall Conditioning (ICWC) is one of the most promising options and has been the subject of considerable recent study on current tokamaks. This paper reports on the findings of such studies performed on European tokamaks, covering a range of plasma-facing materials: TORE SUPRA, TEXTOR, ASDEX Upgrade and JET.
    Journal of Nuclear Materials 12/2010; 415:S1021–S1028. · 2.02 Impact Factor

Publication Stats

92 Citations
67.07 Total Impact Points

Institutions

  • 2014
    • FOM Institute AMOLF
      Amsterdamo, North Holland, Netherlands
  • 2010
    • National Center for Scientific Research Demokritos
      Athínai, Attica, Greece
  • 2005–2008
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Garching bei München, Bavaria, Germany
  • 2007
    • Humboldt-Universität zu Berlin
      • Department of Physics
      Berlin, Land Berlin, Germany
  • 2006
    • Uppsala University
      • Department of Physics and Astronomy
      Uppsala, Uppsala, Sweden