A. Kallenbach

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

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Publications (291)407.48 Total impact

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    ABSTRACT: ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.
    Physics of Plasmas 02/2015; 22:021806. · 2.25 Impact Factor
  • Journal of Nuclear Materials 01/2015; · 2.02 Impact Factor
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    ABSTRACT: The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favorable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. In present tokamaks, this H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density.In gas ramp discharges at the fully tungsten covered ASDEX Upgrade tokamak (AUG), four distinct operational phases are identified in the approach towards the HDL. These phases are a stable H-mode, a degrading H-mode, the breakdown of the H-mode and an L-mode. They are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analyzed. During the evolution, energy losses are increased and a fueling limit is encountered. The latter is correlated to a plateau of electron density in the scrape-off layer (SOL). The well-known extension of the good confinement at high density with high triangularity is reflected in this scheme by extending the first phase to higher densities.In this work, two mechanisms are proposed, which can explain the experimental observations. The fueling limit is most likely correlated to an outward shift of the ionization profile. The additional energy loss channel is presumably linked to a regime of increased radial filament transport in the SOL. The SOL and divertor plasmas play a key role for both mechanisms, in line with the previous hypothesis that the HDL is edge-determined.The four phases are also observed in carbon covered AUG, although the HDL density exhibits a different dependency on the heating power and plasma current. This can be attributed to a changed energy loss channel in the presented scheme.
    Plasma Physics and Controlled Fusion 12/2014; 57(1). · 2.39 Impact Factor
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    ABSTRACT: Operation of DEMO in comparison to ITER will be significantly more demanding, as various additional limitations of physical and technical nature have to be respected. In particular a set of extremely restrictive boundary conditions on divertor operation during and in between ELMs will have to be respected. It is of high importance to describe these limitations in order to consider them as early as possible in the ongoing development of the DEMO concept design. This paper extrapolates the existing physics basis on power and particle exhaust to DEMO.
    Nuclear Fusion 11/2014; 54(11):114003. · 3.24 Impact Factor
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    ABSTRACT: In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales.
    The Review of scientific instruments 03/2014; 85(3):033503. · 1.58 Impact Factor
  • Plasma Physics and Controlled Fusion 12/2013; 55(12):124036. · 2.39 Impact Factor
  • Fusion Engineering and Design 10/2013; 88(9-10):1541-1545. · 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. · 3.24 Impact Factor
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    ABSTRACT: The SOL power decay length (λqλq) deduced from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade with carbon plasma facing components, are presented. Interpretation of the target heat load profiles is performed by using a 1D-fit function which disentangles the upstream λqλq and an effective diffusion in the divertor (S), the latter essentially acting as a power spreading parameter in the divertor volume. It is shown that the so called integral decay length λintλint is approximately given by λint≈λq+1.64×Sλint≈λq+1.64×S. An empirical scaling reveals parametric dependency λq/mm≃0.9·BT-0.7qcyl1.2PSOL0Rgeo0 for type-I ELMy H-modes. Extrapolation to ITER gives λq≃λq≃1 mm. Recent measurements in JET-ILW and from ASDEX Upgrade full-W confirm the results. It is shown that a regression for the divertor power spreading parameter S is not yet possible due to the large effect of different divertor geometries of JET and ASDEX Upgrade Divertor-I and Divertor-IIb.
    Journal of Nuclear Materials 07/2013; 438:S72–S77. · 2.02 Impact Factor
  • AIP Conference Proceedings 06/2013; 1580:271.
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    ABSTRACT: Experiments have been carried out in the TEXTOR, ASDEX Upgrade (AUG) and Alcator C-Mod (C-Mod) tokamaks to study melt-layer motion, macroscopic W-erosion from the melt as well as the changes of material properties such as grain-size and voids. In addition the effect of multiple exposures is studied to judge the potential amelioration of inflicted melt damage. The parallel heat flux at the radial position of the PFCs in the plasma ranges from around q∥ ∼ 45 MW/m2 at TEXTOR up to q∥ ∼ 500 MW/m2 at C-Mod which covers scenarios close to ITER parameters, allowing samples to be exposed and molten even at shallow divertor angles. Melt-layer motion perpendicular to the magnetic field is observed consistent with a Lorentz-force originating from thermoelectric emission of the hot sample. While melting in the limiter geometry at TEXTOR is rather quiescent causing no severe impact on plasma operation, exposure in the divertors of AUG and C-Mod shows significant impact on operation, leading to subsequent disruptions. The power-handling capabilities are severely degraded by forming exposed hill structures and changing the material structure by re-solidifying and re-crystallizing the original material. Melting of W seems highly unfavorable and needs to be avoided especially in light of uncontrolled transients and misaligned PFCs.
    Journal of Nuclear Materials 01/2013; · 2.02 Impact Factor
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    ABSTRACT: Sustained burning plasma operation is the primary target of thermonuclear fusion research and development. To generate economically viable electricity from nuclear fusion requires a high reaction rate over long times with high reliability. Plasma performance must continue to improve above the current levels to reach this target. Non-linear stability limits and internal couplings make optimization of plasma performance a complex, convoluted task, which can be mastered only by assistance of automated control tools. The ASDEX Upgrade fusion experiment has a long tradition in the exploration of physics relations and the subsequent derivation of control strategies. Based on examples from current projects, radiative cooling, NTM stabilization and ELM mitigation, we show typical facets of performance control and how they are implemented by the ASDEX Upgrade Discharge Control System (DCS). Finally, we explain how the methods could be further developed and which additional features would be necessary in the ITER context.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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: Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.
    10/2012;
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    ABSTRACT: ASDEX Upgrade (AUG) is an ITER-shaped divertor tokamak with versatile heating, fueling, exhaust, and control systems. All plasma-facing components (PFCs) are coated with tungsten layers. Plasma scenarios have been adopted that avoid central tungsten accumulation, which can lead to an H–L transition due to excessive central radiative losses. Compared to a carbon-PFC tokamak, the AUG operational space is slightly more weighted toward higher densities and collisionalities. Actual and future planned extensions aim toward reducing the core collisionality while maintaining good power and particle exhaust. These extensions include a solid tungsten outer divertor target, improved pumping, higher ECRH power, and modified ICRF antennas that reduce tungsten sources. The newest element for advanced plasma control is the first set of eight magnetic perturbation coils, which already achieved type-I edge localized mode mitigation in multiple plasma scenarios. Another eight coils have been installed in autumn 2011, allowing the production of mode spectra with $n > 2$ . In parallel to the improved actuator set, an increasing number of diagnostics are brought into real-time state, allowing versatile profile and stability control.
    IEEE Transactions on Plasma Science 03/2012; 40(3):605-613. · 0.95 Impact Factor
  • 20th International Conference on Plasma Surface Interaction in Fusion Devices, to be published in Journal of Nuclear Materials. 01/2012;
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    ABSTRACT: Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen for radiative cooling, discharges with additional heating power up to 23 MW have been achieved, providing good confinement (H98y2 = 1), divertor power loads around 5 MW m−2 and divertor temperatures below 10 eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4 MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters.
    20th International Conference on Plasma Surface Interaction in Fusion Devices, to be published in Journal of Nuclear Materials. 01/2012;
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    ABSTRACT: Tungsten rods of 1×1×3 mm3 were exposed in single H-mode discharges at the outer divertor target plate of ASDEX Upgrade using the divertor manipulator system. Melting of the W rod at a pre-defined time was induced by moving the initially far away outer strike point close to the W-rod position. Visible light emissions of both the W pin and consecutively ejected W droplets were recorded by two fast cameras with crossed viewing cones. The time evolution of the local W source at the pin location was measured by spectroscopic observation of the WI line emission at 400.9 nm and compared to the subsequent increase of tungsten concentration in the confined plasma derived from tungsten vacuum UV line emission. Combining these measurements with the total amount of released tungsten due to the pin melt events and ejected droplets allowed us to derive an estimate of the screening factor for this type of tungsten source. The resulting values of the tungsten divertor retention in the range 10–20 agree with those found in previous studies using a W source of sublimated W(CO)6 vapour at the same exposure location. Ejected droplets were found to be always accelerated in the general direction of the plasma flow, attributed to friction forces and to rocket forces. Furthermore, the vertically inclined target plates cause the droplets, which are repelled by the target plate surface potential due to their electric charge, to move upwards against gravity due to the centrifugal force component parallel to the target plate.
    Physica Scripta 12/2011; · 1.30 Impact Factor
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    ABSTRACT: ELM-resolved divertor target power load studies were conducted for a wide range of discharge conditions in the JET tokamak. The magnetic configuration of these discharges was optimized for the fast divertor IR camera observing the outboard target. It is found that the ELM size estimated from the diamagnetic energy is not uniquely determining the ELM energy load at the divertor target. ELM mid-plane integral deposited power widths between 7 and 18 mm are observed, the inter-ELM widths lie in the range 2.5–6 mm. This ELM broadening is found to widen with ELM size. The temporal evolution of the ELM shape was characterized by rise and decay times. The ELM rise times are found to be in the range expected for ITER (250 µs), but the ELM decay is usually larger than assumed for the ITER design.
    Nuclear Fusion 12/2011; · 3.24 Impact Factor
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    ABSTRACT: Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.
    Plasma Physics and Controlled Fusion 11/2011; 53(12):124017. · 2.39 Impact Factor