M. Tsalas

Culham Centre for Fusion Energy, Abingdon-on-Thames, England, United Kingdom

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Publications (61)99.29 Total impact

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    ABSTRACT: ELMs and Sawteeth, located in different parts of the plasma, are similar from a control engineering point of view. Both manifest themselves through quiescent periods interrupted by periodic collapses. For both, large collapses, following long quiescent periods, have detrimental effects while short periods are associated with decreased confinement. Following the installation of the all metal 'ITER like wall' on JET, sawteeth and ELMs also play an important role by expelling tungsten from the core and edge of the plasma respectively. Control of tungsten has therefore been added to divertor heat load reduction, NTM avoidance and helium ash removal as reasons for requiring ELM and sawtooth control. It is therefore of interest to implement control systems to maintain the sawtooth and ELM frequencies in the desired ranges. On JET, ELM frequency control uses radial field 'kicks' and pellet and gas injection as actuators, while sawtooth control uses ion cyclotron resonance heating (ICRH). JET experiments have, for the first time, established feedback control of the ELM frequency, via real time variation of the injected gas flow [1]. Using this controller in conjunction with pellet injection allows the ELM frequency to be kept as required despite variations in pellet ELM triggering efficiency. JET Sawtooth control experiments have, for the first time, demonstrated that low field side ICRH, as foreseen for ITER, can shorten sawteeth lengthened by central fast ions [2]. The development of ELM and sawtooth control could be key to achieve stable high performance JET discharges with minimal tungsten content. Integrating such schemes into an overall control strategy will be required in future tokamaks and gaining experience on current tokamaks is essential.
    No preview · Article · Jan 2016 · Nuclear Fusion
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    ABSTRACT: The main purpose of this work is to study the dependence of trapped electron modes (TEM) threshold and of electron stiffness on the most relevant plasma parameters. Dedicated transport experiments based on heat flux scans and Te modulation have been performed in JET in TEM dominated plasmas with pure ICRH electron heating and a numerical study using gyrokinetic simulations has been performed with the code GKW. Using multilinear regressions on the experimental data, the stabilizing effect of magnetic shear predicted by theory for our plasma parameters is confirmed while no significant effect of safety factor was found. Good quantitative agreement is found between the TEM thresholds found in the experiments and calculated with linear GKW simulations. Non-linear simulations have given further confirmation of the threshold values and allowed comparison with the values of stiffness found experimentally. Perturbative studies using RF power modulation indicate the existence of an inward convective term for the electron heat flux. Adding NBI power, ion temperature gradient (ITG) modes become dominant and a reduction of with respect to pure ICRH, TEM dominant discharges has been experimentally observed, in spite of increased total electron power. Possible explanations are discussed.
    No preview · Article · Sep 2015 · Nuclear Fusion
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    ABSTRACT: Asymmetrical disruptions may occur during ITER operation and they may be accompanied by large sideways forces and rotation of the asymmetry. This is of particular concern because resonance of the rotating asymmetry with the natural frequencies of the vacuum vessel (and other in-vessel components) could lead to large dynamic amplification of the forces. A significant fraction of non-mitigated JET disruptions have toroidally asymmetric currents that flow partially inside the plasma and partially inside the surrounding vacuum vessel ('wall'). The toroidal asymmetries (otherwise known as the appearance of 3D structures) are clearly visible in the plasma current (Ip) and the first plasma current moments. For the first time we present here the asymmetries in toroidal flux measured by the diamagnetic loops and also propose a physical interpretation. The presented data covers the period of JET operation with a C-wall (JET-C from 2005 until late 2009) and with an ITER-like wall (JET-ILW from 2011 until late 2014), during which pick-up coil and saddle loop data at four toroidally orthogonal locations were routinely recorded. The observed rotations of the Ip asymmetries are in the range from -5 turns to +10 turns (a negative value is counted to the negative plasma current). Initial observations on COMPASS of asymmetric disruptions are presented, which are in line with JET data. The whole of the JET-ILW disruption database and the limited number of COMPASS disruptions examined confirm that the development of the toroidal asymmetry precedes the drop to unity of q95. It is shown that massive gas injection (MGI), which is routinely used to mitigate disruptions, significantly reduces the Ip asymmetries in JET. However, MGI produces fast plasma current quench and consequently high vessel eddy currents, which expose the machine to additional stresses. The effect of the large gas quantity used during the injection is of particular concern as well.
    Full-text · Article · Sep 2015 · Nuclear Fusion
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    ABSTRACT: The influence of the ITER-like Wall (ILW) with divertor target plate made of tungsten (W), on plasma performance in JET H-mode is being investigated since 2011 (see F. Romanelli and references therein). One of the key issues in discharges with low level of D fuelling is observed accumulation of W in the plasma core, which leads to a reduction in plasma performance. To study the interplay between W sputtering on the target plate, penetration of W through the SOL and edge transport barrier (ETB) and its further accumulation in plasma core predictive modelling was launched using a coupled 1.5D core and 2D SOL code JINTRAC (Romanelli, 2014; Cenacchi and Taroni, 1988; Taroni et al., 1992; Wiesen et al., 2006). Simulations reveal the important role of ELMs in W sputtering and plasma density control. Analyses also confirm pivotal role played by the neo-classical pinch of heavy impurities within the ETB.
    No preview · Article · Aug 2015 · Journal of Nuclear Materials
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    ABSTRACT: This paper describes the first development and implementation of a closed loop edge localized mode (ELM) frequency controller using gas injection as the actuator. The controller has been extensively used in recent experiments on JET and it has proved to work well at ELM frequencies in the 15–40 Hz range. The controller responds effectively to a variety of disturbances, generally recovering the requested ELM frequency within approximately 500 ms. Controlling the ELM frequency has become of prime importance in the new JET configuration with all metal walls, where insufficient ELM frequency is associated with excessive tungsten influx. The controller has allowed successful operation near the minimum acceptable ELM frequency where the best plasma confinement can be achieved. Use of the ELM frequency controller in conjunction with pellet injection has enabled investigations of ELM triggering by pellets while maintaining the desired ELM frequency even when pellets fail to trigger ELMs.
    No preview · Article · Jun 2015 · Nuclear Fusion
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    ABSTRACT: New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012 Nat. Commun. 3 624) high field side resonance sawtooth control experiments undertaken at JET, it is found that the sawteeth of L-mode plasmas can be controlled with less accurate alignment between the resonance layer and the sawtooth inversion radius. This advantage, as well as the discovery that sawteeth can be shortened with various antenna phasings, including dipole, indicates that ICRH is a particularly effective and versatile tool that can be used in future fusion machines for controlling sawteeth. Without sawtooth control, neoclassical tearing modes (NTMs) and locked modes were triggered at very low normalised beta. High power H-mode experiments show the extent to which ICRH can be tuned to control sawteeth and NTMs while simultaneously providing effective electron heating with improved flushing of high Z core impurities. Dedicated ICRH simulations using SELFO, SCENIC and EVE, including wide drift orbit effects, explain why sawtooth control is effective with various antenna phasings and show that the sawtooth control mechanism cannot be explained by enhancement of the magnetic shear. Hybrid kinetic-magnetohydrodynamic stability calculations using MISHKA and HAGIS unravel the optimal sawtooth control regimes in these ITER relevant plasma conditions.
    No preview · Article · Dec 2014 · Plasma Physics and Controlled Fusion
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    ABSTRACT: A set of discharges performed with the JET ITER-like wall is investigated with respect to control capabilities on tungsten sources and transport. In attached divertor regimes, increasing fueling by gas puff results in higher divertor recycling ion flux, lower divertor tungsten source, higher ELM frequency and lower core plasma radiation, dominated by tungsten ions. Both pedestal flushing by ELMs and divertor screening (including redeposition) are possibly responsible. For specific scenarios, kicks in plasma vertical position can be employed to increase the ELM frequency, which results in slightly lower core radiation. The application of ion cyclotron radio frequency heating at the very center of the plasma is efficient to increase the core electron temperature gradient and flatten electron density profile, resulting in a significantly lower central tungsten peaking. Beryllium evaporation in the main chamber did not reduce the local divertor tungsten source whereas core radiation was reduced by approximately 50%.
    Full-text · Article · Dec 2014 · Journal of Nuclear Materials
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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.
    Full-text · Article · Apr 2014 · Physics of Plasmas
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    ABSTRACT: New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012 Nat. Commun. 3 624) high field side resonance sawtooth control experiments undertaken at JET, it is found that the sawteeth of L-mode plasmas can be controlled with less accurate alignment between the resonance layer and the sawtooth inversion radius. This advantage, as well as the discovery that sawteeth can be shortened with various antenna phasings, including dipole, indicates that ICRH is a particularly effective and versatile tool that can be used in future fusion machines for controlling sawteeth. Without sawtooth control, neoclassical tearing modes (NTMs) and locked modes were triggered at very low normalised beta. High power H-mode experiments show the extent to which ICRH can be tuned to control sawteeth and NTMs while simultaneously providing effective electron heating with improved flushing of high Z core impurities. Dedicated ICRH simulations using SELFO, SCENIC and EVE, including wide drift orbit effects, explain why sawtooth control is effective with various antenna phasings and show that the sawtooth control mechanism cannot be explained by enhancement of the magnetic shear. Hybrid kinetic-magnetohydrodynamic stability calculations using MISHKA and HAGIS unravel the optimal sawtooth control regimes in these ITER relevant plasma conditions.
    No preview · Article · Jan 2014
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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.
    Full-text · Article · Sep 2013 · Nuclear Fusion
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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.
    Full-text · Article · Aug 2013 · Journal of Nuclear Materials
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    ABSTRACT: The presence of dust particles in a fusion plasma is recognized as a serious issue for safe and efficient operation of the ITER tokamak. The paper presents an in situ laser assisted method for characterization of dust from thermal emission of the particles. The method was developed in the TEXTOR tokamak with the use of Thomson scattering (TS). The diagnostic is capable to detect single particles and measure the dust density profile along the laser probing axis, velocity distribution of dust particles along this axis as well as surface temperature and size of the detected particles.
    No preview · Article · Jul 2013 · Journal of Nuclear Materials
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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.
    No preview · Article · Apr 2013 · Plasma Physics and Controlled Fusion
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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.
    No preview · Article · Jan 2013 · Plasma Physics and Controlled Fusion
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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.
    Full-text · Article · Nov 2012 · Plasma Physics and Controlled Fusion
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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.
    No preview · Conference Paper · Oct 2012
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    ABSTRACT: A new technique has been developed to produce plasmas with improved confinement relative to the H-98,H-y2 scaling law (ITER Physics Expert Groups on Confinement and Transport and Confinement Modelling and Database ITER Physics Basics Editors and ITER EDA 1999 Nucl. Fusion 39 2175) on the JET tokamak. In the mid-size tokamaks ASDEX upgrade and DIII-D heating during the current formation is used to produce a flat q-profile with a minimum close to 1. On JET this technique leads to q-profiles with similar minimum q but opposite to the other tokamaks not to an improved confinement state. By changing the method utilizing a faster current ramp with temporary higher current than in the flattop (current overshoot) plasmas with improved confinement (H-98,H-y2 = 1.35) and good stability (beta(N) approximate to 3) have been produced and extended to many confinement times only limited by technical constraints. The increase in H-98,H-y2-factor is stronger with more heating power as can be seen in a power scan. The q-profile development during the high power phase in JET is reproduced by current diffusion calculated by TRANSP and CRONOS. Therefore the modifications produced by the current overshoot disappear quickly from the edge but the confinement improvement lasts longer, in some cases up to the end of the heating phase.
    Full-text · Article · Sep 2012 · Plasma Physics and Controlled Fusion
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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.
    No preview · Article · Jun 2012 · Plasma Physics and Controlled Fusion
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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.
    Full-text · Article · Mar 2012 · Journal of Instrumentation
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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.
    No preview · Article · Nov 2011 · Plasma Physics and Controlled Fusion

Publication Stats

439 Citations
99.29 Total Impact Points

Institutions

  • 2015
    • Culham Centre for Fusion Energy
      Abingdon-on-Thames, England, United Kingdom
  • 2014-2015
    • FOM Institute AMOLF
      Amsterdamo, North Holland, Netherlands
  • 2009-2014
    • National Center for Scientific Research Demokritos
      Athínai, Attica, Greece
  • 2013
    • Dutch Institute for Fundamental Energy Research
      Nieuwegen, Utrecht, Netherlands
  • 2011
    • National Technical University of Athens
      Athínai, Attica, Greece
  • 2007-2008
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Garching bei München, Bavaria, Germany