G. Pautasso

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

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Publications (102)153 Total impact

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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.
    Nuclear Fusion 06/2015; 55(6):063030. DOI:10.1088/0029-5515/55/6/063030 · 3.24 Impact Factor
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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.
    Nuclear Fusion 03/2015; 55(3):033015. DOI:10.1088/0029-5515/55/3/033015 · 3.24 Impact Factor
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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.
    Physics of Plasmas 01/2015; 22(2). DOI:10.1063/1.4901251 · 2.25 Impact Factor
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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.
    41st EPS conference on Plasma Physics, Berlin, Germany; 01/2014
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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.
    41st EPS conference on Plasma Physics, Berlin, Germany; 01/2014
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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.
    Fusion Engineering and Design 10/2013; 88(6-8):1297-1301. DOI:10.1016/j.fusengdes.2013.01.103 · 1.15 Impact Factor
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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. DOI:10.1088/0029-5515/53/10/104003 · 3.24 Impact Factor
  • Plasma Physics and Controlled Fusion 07/2013; 55(7):074007. DOI:10.1088/0741-3335/55/7/074007 · 2.39 Impact Factor
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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.
    Nuclear Fusion 02/2013; 53(3):033002. DOI:10.1088/0029-5515/53/3/033002 · 3.24 Impact Factor
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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 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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
    Fusion Engineering and Design 10/2011; 86(9):1915-1919. DOI:10.1016/j.fusengdes.2011.02.096 · 1.15 Impact Factor
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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.
    Fusion Engineering and Design 10/2011; 86(6):1039-1044. DOI:10.1016/j.fusengdes.2011.01.069 · 1.15 Impact Factor
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    ABSTRACT: Knowledge discovery consists of finding new knowledge from databases where dimension, complexity, or amount of data is prohibitively large for human observation alone. The need for efficient data visualization and clustering is often faced, for instance, in the analysis, monitoring, fault detection, or prediction of various engineering plants. In this paper, two clustering techniques, K-means and Self-Organizing Maps, are used for the identification of characteristic regions for plasma scenario in nuclear fusion experimental devices. The choice of the number of clusters, which heavily affects the performance of the mapping, is firstly faced. Then, the ASDEX Upgrade Tokamak high-dimensional operational space is mapped into lower-dimensional maps, allowing to detect the regions with high risk of disruption, and, finally, the current process state and its history in time are visualized as a trajectory on the Self-Organizing Map, in order to predict the safe or disruptive state of the plasma. KeywordsKnowledge discovery–Clustering–Self-Organizing Maps–Tokamak–Disruptions
    Neural Computing and Applications 09/2011; 20(6):851-863. DOI:10.1007/s00521-011-0529-2 · 1.76 Impact Factor
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    ABSTRACT: The mapping of the n-dimensional plasma parameter space of ASDEX Upgrade has been performed using a 2-dimensional Self Organizing Map, which reveals the map potentiality in data visualization. The proposed approach allows us the definition of simple displays capable of presenting meaningful information on the actual state of the plasma, but it also suggests to use the Self Organizing Map as a disruption predictor. In this paper, different criteria have been studied to associate the risk of disruption of each cluster in the map to a disruption alarm threshold. Data for this study comes from ASDEX Upgrade experiments executed between July 2002 and November 2009. The prediction performance of the proposed system has been evaluated on a set of discharges different from those used for the map training, obtaining a quite good prediction success rate. A deep analysis of the wrong predictions has been performed in order to identify possible common causes, and some criteria to increase prediction performance have been derived.
    Fusion Engineering (SOFE), 2011 IEEE/NPSS 24th Symposium on; 07/2011
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    ABSTRACT: In this paper, a set of simple predictive criteria, each optimized for a given type of disruption, is explored. Disruptions that occurred in the years from 2005 to 2009 in the ASDEX Upgrade tokamak are classified into several types in a first step. Then, discriminant analysis is used as the main approach to the disruption prediction and a log-linear discriminant function, constructed with five global plasma parameters that have been selected from an initial group of ten variables, is derived for the edge cooling disruptions. The function is tested off-line over 308 discharges and is shown to work reliably. It describes a clear dependence of the disruption boundary on the plasma parameters.
    Nuclear Fusion 06/2011; 51(6). DOI:10.1088/0029-5515/51/6/063039 · 3.24 Impact Factor
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    ABSTRACT: Due to the complexity of the phenomena involved, a self-consistent physical model for the prediction of the halo current is not available. Therefore the ITER specifications of the spatial distribution and evolution of the halo current rely on empirical assumptions. This paper presents the results of an extensive analysis of the halo current measured in ASDEX Upgrade with particular emphasis on the evolution of the halo region, on the magnitude and time history of the halo current, and on the structure and duration of its toroidal and poloidal asymmetries. The effective length of the poloidal path of the halo current in the vessel is found to be rather insensitive to plasma parameters. Large values of the toroidally averaged halo current are observed in both vertical displacement events and centred disruptions but last a small fraction of the current quench; they coincide typically with a large but short-lived MHD event.
    Nuclear Fusion 04/2011; 51(4). DOI:10.1088/0029-5515/51/4/043010 · 3.24 Impact Factor
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    ABSTRACT: The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20MW heating power at Ip up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above ne = 1.2×1020 m−3 and low q95 operation at Bt = 1.8 T. Central O2-mode heatingwas successfully used in high P/R discharges with 20MWtotal heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the Te/Ti ratio could be well reproduced. Doppler reflectometry studies at the L–H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative Er well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of ne and Te. The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfv´en eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms.
    Nuclear Fusion 01/2011; 51(8):094012. DOI:10.1088/0029-5515/51/9/094012 · 3.24 Impact Factor
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    ABSTRACT: Plasma disruptions represent a hazard for the structural integrity of ITER. The contribution of the existing tokamaks to this project consists in refining the characterization of the disruption loads and their extrapolation, on the basis of physical models, and in learning to predict and mitigate disruptions. The ASDEX Upgrade research program covers these specific topics and this contribution reports on significant progress made in these areas. (I) The formation and evolution of the halo region is analyzed with MHD-transport codes and extrapolation to ITER is discussed. (II) Discriminant analysis is applied to each disruption group in order to determine the most significant plasma variables, which allow for classification, and is being used to discern between stable and pre-disruptive plasma states. (III) Progress has been made with MGI in attaining an effective electron density equivalent to 24 % of the critical one (necessary for the collisional suppression of runaway electrons) and in studying the redistribution of the injected gas in the plasma.
    Nuclear Fusion 11/2010; 51(10). DOI:10.1088/0029-5515/51/10/103009 · 3.24 Impact Factor
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    ABSTRACT: In this paper, a neural predictor has been built using plasma discharges selected from two years of ASDEX Upgrade experiments, from July 2002 to July 2004. In order to test the real-time prediction capability of the system, its performance has been evaluated using discharges coming from different experimental campaigns, from June 2005 to July 2007. All disruptions that occurred in the chosen experimental campaigns were included with the exception of those occurring in the ramp-up phase, in the ramp-down phase (if the disruption does not happen in the first 100 ms), those caused by massive gas injection and disruptions following vertical displacement events. The large majority of selected disruptions are of the cooling edge type and typically preceded by the growth of tearing modes, degradation of the thermal confinement and enhanced plasma radiation. A very small percentage of them happen at large beta after a short precursor phase. For each discharge, seven plasma diagnostic signals have been selected from numerous signals available in real-time. During the training procedure, a self-organizing map has been used to reduce the database size in order to improve the training of the neural network. Moreover, an optimization procedure has been performed to discriminate between safe and pre-disruptive phases. The prediction success rate has been further improved, performing an adaptive training of the network whenever a missed alarm is triggered by the predictor.
    Nuclear Fusion 06/2010; 50(7):075004. DOI:10.1088/0029-5515/50/7/075004 · 3.24 Impact Factor
  • The Review of scientific instruments 03/2010; 81(3):039903. DOI:10.1063/1.3340955 · 1.58 Impact Factor

Publication Stats

1k Citations
153.00 Total Impact Points

Institutions

  • 1997–2015
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 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