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Author

V. Igochine

Bio: V. Igochine is an academic researcher from Max Planck Society. The author has contributed to research in topics: ASDEX Upgrade & Tokamak. The author has an hindex of 33, co-authored 154 publications receiving 3165 citations.


Papers
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Journal ArticleDOI
TL;DR: In this paper, the evolution of a reversed-field-pinch plasma towards a self-organized single-helicity state suggests that instability problems, which have previously hindered the development of these devices, could now be overcome.
Abstract: A reversed-field pinch is a toroidal device for magnetically confining plasmas, and a potential alternative to the tokamak for a future fusion reactor. Observations of the evolution of a reversed-field-pinch plasma towards a self-organized single-helicity state suggest that instability problems, which have previously hindered the development of these devices, could now be overcome.

260 citations

Journal ArticleDOI
A. C. C. Sips1, R. Arslanbekov1, C. Atanasiu2, W. Becker1, G. Becker1, K. Behler1, K.H. Behringer1, A. Bergmann1, R. Bilato1, D. Bolshukhin1, K. Borrass1, Bastiaan J. Braams2, Marco Brambilla1, F. Braun1, A. Buhler1, Garrard Conway1, D. P. Coster1, R. Drube1, R. Dux1, S.M. Egorov1, T. Eich1, K. Engelhardt1, H.-U. Fahrbach1, Ursel Fantz1, Helmut Faugel1, M. Foley2, K. B. Fournier2, P. Franzen1, Julia Fuchs1, J. Gafert1, G. Gantenbein2, O. Gehre1, A. Geier1, J. Gernhardt1, O. Gruber1, A. Gude1, Sibylle Günter1, G. Haas1, Dirk Hartmann1, B. Heger3, Bernd Heinemann1, Albrecht Herrmann1, J. Hobirk1, F. Hofmeister1, H. Hohenöcker1, L. D. Horton1, V. Igochine1, D. Jacobi1, Martin Jakobi1, Frank Jenko1, A. Kallenbach1, O. J. W. F. Kardaun1, Michael Kaufmann1, A. Keller1, Alexander Kendl1, Junghee Kim1, K. K. Kirov1, R. Kochergov1, H. Kollotzek1, W. Kraus1, K. Krieger1, B. Kurzan1, Peter Lang1, Philipp Lauber1, Martin Laux1, F. Leuterer1, A. Lohs1, A. Lorenz1, C. F. Maggi1, H. Maier1, K. Mank1, M. E. Manso2, M. Maraschek1, K. F. Mast1, Patrick J. McCarthy1, D. Meisel1, H. Meister1, Fernando Meo1, R. Merkel1, D. Merkl1, V. Mertens1, F. Monaco1, A. Mück1, H. W. Müller1, M. Münich1, H. D. Murmann1, Y.-S. Na1, G. Neu1, R. Neu1, J. Neuhauser1, J.-M. Noterdaeme1, I. Nunes1, G. Pautasso1, A. G. Peeters1, G. V. Pereverzev1, S. D. Pinches1, Emanuele Poli1, M. Proschek2, R. Pugno1, E. Quigley1, Gerhard Raupp1, T. Ribeiro4, R. Riedl1, S. Riondato1, V. Rohde1, J. Roth1, F. Ryter1, S. Saarelma2, W. Sandmann1, S. Schade1, H. B. Schilling1, Wolf-Dieter Schneider1, G. Schramm1, S. Schweizer1, Bruce D. Scott1, U. Seidel1, F. Serra2, S. Sesnic1, C. Sihler1, Ana Elisa Bauer de Camargo Silva2, E. Speth1, A. Stäbler1, K.-H. Steuer1, J. Stober1, B. Streibl1, E. Strumberger1, W. Suttrop1, A. Tabasso1, A. Tanga1, G. Tardini1, C. Tichmann1, W. Treutterer1, M. Troppmann1, P. Varela2, O. Vollmer1, D. Wagner1, U. Wenzel1, F. Wesner1, Robert Wolf1, E. Wolfrum1, E. Würsching1, Q. Yu1, D. Zasche1, Thomas Zehetbauer1, H. P. Zehrfeld1, H. Zohm1 
TL;DR: In this paper, a modified divertor configuration for ASDEX upgrade, allowing operation at higher triangularity, and with a changed neutral beam injection (NBI) system, for a more tangential, off-axis beam deposition.
Abstract: Recent experiments at ASDEX Upgrade have achieved advanced scenarios with high β N (>3) and confinement enhancement over ITER98(y, 2) scaling, H H98y2 = 1.1-1.5, in steady state. These discharges have been obtained in a modified divertor configuration for ASDEX Upgrade, allowing operation at higher triangularity, and with a changed neutral beam injection (NBI) system, for a more tangential, off-axis beam deposition. The figure of merit, β N H ITER89-P , reaches up to 7.5 for several seconds in plasmas approaching stationary conditions. These advanced tokamak discharges have low magnetic shear in the centre, with q on-axis near 1, and edge safety factor, q 95 in the range 3.3-4.5. This q-profile is sustained by the bootstrap current, NBI-driven current and fishbone activity in the core. The off-axis heating leads to a strong peaking of the density profile and impurity accumulation in the core. This can be avoided by adding some central heating from ion cyclotron resonance heating or electron cyclotron resonance heating, since the temperature profiles are stiff in this advanced scenario (no internal transport barrier). Using a combination of NBI and gas fuelling line, average densities up to 80-90% of the Greenwald density are achieved, maintaining good confinement. The best integrated results in terms of confinement, stability and ability to operate at high density are obtained in highly shaped configurations, near double null, with δ = 0.43. At the highest densities, a strong reduction of the edge localized mode activity similar to type II activity is observed, providing a steady power load on the divertor, in the range of 6 MW m -2 , despite the high input power used (> 10 MW).

111 citations

Journal ArticleDOI
Rudolf Neu1, M. Balden1, V. Bobkov1, R. Dux1  +153 moreInstitutions (6)
TL;DR: In this article, it was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction and after the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-mode confinement with H factors close to one was achieved.
Abstract: ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.

103 citations

Journal Article
R. Neu1, M. Balden1, V. Bobkov1, R. Dux1  +153 moreInstitutions (2)
TL;DR: In this article, it was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction and after the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-mode confinement with H factors close to one was achieved.
Abstract: ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.

101 citations

Journal ArticleDOI
H. Zohm1, C. Angioni1, R. Arslanbekov1, C. V. Atanasiu2  +164 moreInstitutions (9)
TL;DR: In this paper, the authors studied the power-decay length of the divertor tokamak ASDEX upgrade and showed that the drift-wave dynamics in the inner divertor and the outer divertor can be modelled by a gyro-fluid code and point to the dominance of drift waves.
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 x 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.

90 citations


Cited by
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Journal ArticleDOI
01 Apr 1988-Nature
TL;DR: In this paper, a sedimentological core and petrographic characterisation of samples from eleven boreholes from the Lower Carboniferous of Bowland Basin (Northwest England) is presented.
Abstract: Deposits of clastic carbonate-dominated (calciclastic) sedimentary slope systems in the rock record have been identified mostly as linearly-consistent carbonate apron deposits, even though most ancient clastic carbonate slope deposits fit the submarine fan systems better. Calciclastic submarine fans are consequently rarely described and are poorly understood. Subsequently, very little is known especially in mud-dominated calciclastic submarine fan systems. Presented in this study are a sedimentological core and petrographic characterisation of samples from eleven boreholes from the Lower Carboniferous of Bowland Basin (Northwest England) that reveals a >250 m thick calciturbidite complex deposited in a calciclastic submarine fan setting. Seven facies are recognised from core and thin section characterisation and are grouped into three carbonate turbidite sequences. They include: 1) Calciturbidites, comprising mostly of highto low-density, wavy-laminated bioclast-rich facies; 2) low-density densite mudstones which are characterised by planar laminated and unlaminated muddominated facies; and 3) Calcidebrites which are muddy or hyper-concentrated debrisflow deposits occurring as poorly-sorted, chaotic, mud-supported floatstones. These

9,929 citations

Journal ArticleDOI
TL;DR: A review of recent advances in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed in this paper.
Abstract: Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

1,051 citations

Journal ArticleDOI
TL;DR: The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions.
Abstract: The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

798 citations

Book
19 Dec 2003
TL;DR: In this article, the Equations of Gas Dynamics and Magnetoplasmas Dynamics were studied, as well as Magnetoplasma Stability and Transport in Magnetplasmas and Magnetic Stability.
Abstract: 1 The Equations of Gas Dynamics 2 Magnetoplasma Dynamics 3 Waves in Magnetoplasmas 4 Magnetoplasma Stability 5 Transport in Magnetoplasmas 6 Extensions of Theory Bibliography Index

748 citations

Journal Article
TL;DR: The advantages of nuclear fusion as an energy source and research progress in this area are summarized in this article, where the current state of the art is described, including the Compact Ignition Tokamak (CIT), International Thermonuclear Experimental Reactor (ITER), and a US design called TIBER II.
Abstract: The advantages of nuclear fusion as an energy source and research progress in this area are summarized. The current state of the art is described. Laser fusion, inertial confinement fusion, and magnetic fusion (the tokamak) are explained, the latter in some detail. Remaining problems and planned future reactors are considered. They are the Compact Ignition Tokamak (CIT), the International Thermonuclear Experimental Reactor (ITER), and a US design called TIBER II. The design of the latter is shown. >

596 citations