G. Pautasso

Max Planck Institute for Plasma Physics, Arching, Bavaria, Germany

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Publications (111)154.54 Total impact

  • E. Fable · G. Pautasso · M. Lehnen · R. Dux · M. Bernert · A. Mlynek

    No preview · Article · Feb 2016 · Nuclear Fusion
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    ABSTRACT: The amplitude of locked instabilities, likely magnetic islands, seen as precursors to disruptions has been studied using data from the JET, ASDEX Upgrade and COMPASS tokamaks. It was found that the thermal quench, that often initiates the disruption, is triggered when the amplitude has reached a distinct level. This information can be used to determine thresholds for simple disruption prediction schemes. The measured amplitude in part depends on the distance of the perturbation to the measurement coils. Hence the threshold for the measured amplitude depends on the mode location (i.e. the rational q-surface) and thus indirectly on parameters such as the edge safety factor, q95, and the internal inductance, li(3), that determine the shape of the q-profile. These dependencies can be used to set the disruption thresholds more precisely. For the ITER baseline scenario, with typically q95=3.2, li(3)=0.9 and taking into account the position of the measurement coils on ITER , the maximum allowable measured locked mode amplitude normalized to engineering parameters was estimated to be a·BML(rc)/Ip = 0.92 m mT/MA, or directly as a fraction edge poloidal magnetic field: BML(rc)/Bq(a) = 5·10-3. But these values decrease for operation at higher q95 or lower li(3). The analysis found furthermore that the above empirical criterion to trigger a thermal quench is more consistent with a criterion derived with the concept of a critical island size, i.e. the thermal quench seemed to be triggered at a distinct island width.
    Full-text · Article · Feb 2016 · Nuclear Fusion
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    ABSTRACT: The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.
    No preview · Article · Aug 2015 · Journal of Nuclear Materials
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    ABSTRACT: A multi-device database of disruption characteristics has been developed under the auspices of the International Tokamak Physics Activity magneto-hydrodynamics topical group. The purpose of this ITPA disruption database (IDDB) is to find the commonalities between the disruption and disruption mitigation characteristics in a wide variety of tokamaks in order to elucidate the physics underlying tokamak disruptions and to extrapolate toward much larger devices, such as ITER and future burning plasma devices. In contrast to previous smaller disruption data collation efforts, the IDDB aims to provide significant context for each shot provided, allowing exploration of a wide array of relationships between pre-disruption and disruption parameters. The IDDB presently includes contributions from nine tokamaks, including both conventional aspect ratio and spherical tokamaks. An initial parametric analysis of the available data is presented. This analysis includes current quench rates, halo current fraction and peaking, and the effectiveness of massive impurity injection. The IDDB is publicly available, with instruction for access provided herein.
    Full-text · Article · Jun 2015 · Nuclear Fusion
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    ABSTRACT: Experiments of disruption mitigation with massive gas injection are conducted in ASDEX Upgrade with fast valves located close to the plasma. The valves and the dedicated experiment are described in this paper. The dependence of the fuelling efficiency on plasma and gas parameters is documented and discussed. Several sources of uncertainties affecting its evaluation and physical interpretation have been addressed. An actual fuelling efficiency of 40% has been reached for neon injection with valves close to the plasma and for gas quantities relevant for the thermal and current quench mitigation of ITER. Refuelling the plasma after thermal quench is shown to be feasible; this result opens the possibility of raising the density in a runaway beam and therefore of increasing the collisional drag on and the radiative energy losses of the fast electrons.
    No preview · Article · Mar 2015 · Nuclear Fusion
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    ABSTRACT: An overview of the present status of research toward the final design of the ITER disruption mitigation system (DMS) is given. The ITER DMS is based on massive injection of impurities, in order to radiate the plasma stored energy and mitigate the potentially damaging effects of disruptions. The design of this system will be extremely challenging due to many physics and engineering constraints such as limitations on port access and the amount and species of injected impurities. Additionally, many physics questions relevant to the design of the ITER disruption mitigation system remain unsolved such as the mechanisms for mixing and assimilation of injected impurities during the rapid shutdown and the mechanisms for the subsequent formation and dissipation of runaway electron current.
    Full-text · Article · Jan 2015 · Physics of Plasmas
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    ABSTRACT: ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs in both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically, the Be PFCs of JET led to a very small low-Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady-state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by codeposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N-2) (AUG and JET) or by combining N-2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N-2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fueling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however, in both devices the W content is increased during ICRH operation most probably due to increased W sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20%-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not yet clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N-2 seeding observed in both devices hints to the fact, that low-Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.
    No preview · Article · Mar 2014 · IEEE Transactions on Plasma Science
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    ABSTRACT: A disruption of a tokamak discharge is a sudden loss of confinement, or thermal quench, in turn resulting in a quench of the plasma current. The fast release of thermal and magnetic energy could result in very large thermal and electromagnetic loads on the surrounding structures, such plasma facing components or the vessel, especially in large devices such as JET and ITER. Understandably, considerable research efforts are dedicated to develop both timely detectors of these events and mitigating actions. Magneto-hydrodynamic (MHD) instabilities are often seen as precursors to disruptions. The growth of large, overlapping, magnetic islands is thought to be behind the destruction of the flux surface structure that provides the plasma confinement, triggering the thermal quench [1-4]. The detection of these modes is used to predict disruptions. Usually the analysis of these instabilities focuses on how early and at what level they can first be detected [5]. This paper will investigate a different but related question; is there a specific maximum perturbation level that triggers a thermal quench? This study provides experimental insight in the processes that may trigger tokamak disruptions. The perturbation amplitudes that trigger thermal quenches in JET and ASDEX Upgrade are compared and the results form a strong physics basis to determine protection thresholds to be used at future devices, such as ITER.
    Full-text · Conference Paper · Jan 2014
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    G. Pautasso · P.C. de Vries
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    ABSTRACT: Most of the existing tokamaks implement disruption protection measurements and can initiate a slow or fast (and mitigated) emergency shutdown ; however no reliable disruption prediction system, which is portable to ITER, currently exists in present-day machines. A premise for avoiding or predicting unavoidable disruptions is knowing under which conditions they develop. In this paper, after a short discussion of the disruption rate during the ASDEX Upgrade (AUG) lifetime, the causes of the disruptions that occurred in 2013 (part of the 2012-2013 experimental campaign) are discussed. When possible, disruptions with similar causes are categorized according to the classification system used for JET [1]; in this process, attention has been paid to the chain of precursors preceding the instability. The plasma state directly before the thermal quench (TQ) is discussed. This comparison with JET will provide information on how universal these events are.
    Full-text · Conference Paper · Jan 2014
  • R. Aledda · B. Cannas · A. Fanni · G. Sias · G. Pautasso
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    ABSTRACT: In this paper, a disruption prediction system for ASDEX Upgrade has been proposed that does not require disruption terminated experiments to be implemented. The system consists of a data-based model, which is built using only few input signals coming from successfully terminated pulses. A fault detection and isolation approach has been used, where the prediction is based on the analysis of the residuals of an auto regressive exogenous input model. The prediction performance of the proposed system is encouraging when it is applied to the same set of campaigns used to implement the model. However, the false alarms significantly increase when we tested the system on discharges coming from experimental campaigns temporally far from those used to train the model. This is due to the well know aging effect inherent in the data-based models. The main advantage of the proposed method, with respect to other data-based approaches in literature, is that it does not need data on experiments terminated with a disruption, as it uses a normal operating conditions model. This is a big advantage in the prospective of a prediction system for ITER, where a limited number of disruptions can be allowed.
    No preview · Article · Oct 2013 · Fusion Engineering and Design
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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: The modelling of tokamak scenarios requires the simultaneous solution of both the time evolution of the plasma kinetic profiles and of the magnetic equilibrium. Their dynamical coupling involves additional complications, which are not present when the two physical problems are solved separately. Difficulties arise in maintaining consistency in the time evolution among quantities which appear in both the transport and the Grad-Shafranov equations, specifically the poloidal and toroidal magnetic fluxes as a function of each other and of the geometry. The required consistency can be obtained by means of iteration cycles, which are performed outside the equilibrium code and which can have different convergence properties depending on the chosen numerical scheme. When these external iterations are performed, the stability of the coupled system becomes a concern. In contrast, if these iterations are not performed, the coupled system is numerically stable, but can become physically inconsistent. By employing a novel scheme (Fable E et al 2012 Nucl. Fusion submitted), which ensures stability and physical consistency among the same quantities that appear in both the transport and magnetic equilibrium equations, a newly developed version of the ASTRA transport code (Pereverzev G V et al 1991 IPP Report 5/42), which is coupled to the SPIDER equilibrium code (Ivanov A A et al 2005 32nd EPS Conf. on Plasma Physics (Tarragona, 27 June-1 July) vol 29C (ECA) P-5.063), in both prescribed- and free-boundary modes is presented here for the first time. The ASTRA-SPIDER coupled system is then applied to the specific study of the modelling of controlled current ramp-up in ASDEX Upgrade discharges.
    No preview · Article · Jul 2013 · Plasma Physics and Controlled Fusion
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    ABSTRACT: The coupled system consisting of 1D radial transport equations and the quasi-static 2D magnetic equilibrium equation for axisymmetric systems (tokamaks) is known to be prone to numerical instabilities, either due to propagation of numerical errors in the iteration process, or due to the choice of the numerical scheme itself. In this paper, a possible origin of these instabilities, specifically associated with the latter condition, is discussed and an approach is chosen, which is shown to have good accuracy and stability properties. This scheme is proposed to be used within those codes for which the poloidal flux ψ is the quantity solved for in the current diffusion equation. Mathematical arguments are used to study the convergence properties of the proposed scheme.
    No preview · Article · Feb 2013 · Nuclear Fusion

  • No preview · Article · Jan 2013

  • No preview · Article · Jan 2013
  • Source
    [Show abstract] [Hide abstract]
    ABSTRACT: ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically the Be PFCs of JET led to a very small low-Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by co-deposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N2) (AUG and JET) or by combining N2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fuelling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however in both devices the W content is increased during ICRH operation most probably due to increased W sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not ye- clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N2 seeding observed in both devices hints to the fact, that low-Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.
    Full-text · Conference Paper · Jan 2013
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    ABSTRACT: The Self-Organizing Map is a computational method for the visualization and analysis of high-dimensional data. Self Organizing Maps have been applied to ASDEX Upgrade data to define an ordered mapping of an 8-dimensional plasma parameter space onto a regular, 2-dimensional grid. The map has been used to track the plasma trajectory during the experiments and monitor the disruption risk. In order to face with ever new operational conditions, a periodical updating of the SelfOrganizing Map is proposed.
    No preview · Conference Paper · Jan 2012

  • No preview · Article · Jan 2012
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    ABSTRACT: During asymmetric vertical displacement events (AVDEs) associated with the kink mode of the plasma two asymmetry phenomena were observed in existing tokamaks, in particular in JET [1]. The related halo currents flowing in the passive structure were identified as the cause of asymmetric EM loads on tokamak components. The first phenomenon is a toroidal peak of the poloidal halo current that flows in the passive structure. The second phenomenon is that the toroidal plasma current is not uniform toroidally, so a toroidally non-uniform current flows in the vessel [2]. The specification of the expected characteristics of both phenomena as well as of the consequent asymmetric loads in ITER are summarized here. The related loads are specified for likely, unlikely and extremely unlikely AVDEs.
    Full-text · Article · Oct 2011 · Fusion Engineering and Design
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    B. Cannas · A. Fanni · G. Pautasso · G. Sias
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    ABSTRACT: In this paper, an adaptive neural system has been built to predict the risk of disruption at ASDEX Upgrade. The system contains a Self Organizing Map, which determines the ‘novelty’ of the input of a Multi Layer Perceptron predictor module. The answer of the MLP predictor will be inhibited whenever a novel sample is detected. Furthermore, it is possible that the predictor produces a wrong answer although it is fed with known samples. In this case, a retraining procedure will be performed to update the MLP predictor in an incremental fashion using data coming from both the novelty detection, and from wrong predictions. In particular, a new update is performed whenever a missed alarm is triggered by the predictor.The performance of the adaptive predictor during the more recent experimental campaigns until November 2009 has been evaluated.
    Full-text · Article · Oct 2011 · Fusion Engineering and Design

Publication Stats

2k Citations
154.54 Total Impact Points

Institutions

  • 1994-2015
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 2009
    • Hefei Institute of Physical Sciences, Chinese Academy of Sciences
      Luchow, Anhui Sheng, China
  • 2005
    • Rechenzentrum Garching (RZG) of the Max Planck Society and the IPP
      Arching, Bavaria, Germany
  • 1990
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, New Jersey, United States