A. Kallenbach

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

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Publications (310)472 Total impact

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    Dataset: Talk
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    Dataset: talk EX2-6
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    ABSTRACT: The first stable completely detached H-mode plasma in the full tungsten ASDEX Upgrade has been achieved. Complete detachment of both targets is induced by nitrogen seeding into the divertor. Two new phases are added to the detachment classification described in Potzel et al (2014 Nucl. Fusion 54 013001): first, the line integrated density increases by about 15% with partial detachment of the outer divertor. Second, complete detachment of both targets is correlated to the appearance of intense, strongly localized, stable radiation at the X-point. Radiated power fractions, frad, increase from about 50% to 85% with nitrogen seeding. X-point radiation is accompanied by a loss of pedestal top plasma pressure of about 60%. However, the core pressure at ρpol < 0.7 changes only by about 10%. H98 = 0.8-1.0 is observed during detached operation. With nitrogen seeding the edge-localized mode (ELM) frequency increases from the 100 Hz range to a broadband distribution at 1-2 kHz with a large reduction in ELM size.
    Nuclear Fusion 03/2015; 55(3):033004. DOI:10.1088/0029-5515/55/3/033004 · 3.24 Impact Factor
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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(2):021806. DOI:10.1063/1.4907901 · 2.25 Impact Factor
  • Journal of Nuclear Materials 01/2015; DOI:10.1016/j.jnucmat.2015.01.012 · 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). DOI:10.1088/0741-3335/57/1/014038 · 2.39 Impact Factor
  • Journal of Nuclear Materials 12/2014; DOI:10.1016/j.jnucmat.2014.12.019 · 2.02 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. DOI:10.1088/0029-5515/54/11/114003 · 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. DOI:10.1063/1.4867662 · 1.58 Impact Factor
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    ABSTRACT: The W-transport in the core plasma of JET is investigated experimentally by deriving the W-concentration profiles from the modelling of the signals of the soft x-ray cameras. For the case of pure neutral beam heating W accumulates in the core (r/a < 0.3) approaching W-concentrations of 10(-3) in between the sawtooth crashes, which flatten the W-profile to a concentration of about 3 x 10(-5). When central Ion cyclotron resonant heating is additionally applied the core W-concentration decays in phases that exhibit a changed mode activity, while also the electron temperature increases and the density profile becomes less peaked. The immediate correlation between the change of magnetohydrodymanic (MHD) and the removal of W from the plasma core supports the hypothesis that the change of the MHD activity is the underlying cause for the change of transport. Furthermore, a discharge from the ASDEX Upgrade is investigated. In this case the plasma profiles exhibit small changes only, while the most prominent change occurs in the W-content of the confined plasma caused by the reduction of the fuelling deuterium gas puff. Concomintantly, the W-concentration profiles in the core plasma r/a < 0.2 steepen up reminescent to the well-known connection between central radiation and transport during cases with strong, established W-accumulation, while in the present analysis such a causality between the two during the onset of W-accumulation could not be pinned down. Both case studies exemplify that small changes of the core parameters of a plasma my influence the W-transport in the plasma core drastically.
    Plasma Physics and Controlled Fusion 12/2013; 55(12):124036. DOI:10.1088/0741-3335/55/12/124036 · 2.39 Impact Factor
  • Plasma Physics and Controlled Fusion 12/2013; 55(12):124041. DOI:10.1088/0741-3335/55/12/124041 · 2.39 Impact Factor
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    ABSTRACT: In both JET and ASDEX Upgrade (AUG) the plasma energy confinement has been affected by the presence of a metal wall by the requirement of increased gas fuelling to avoid tungsten pollution of the plasma. In JET with a beryllium/tungsten wall the high triangularity baseline H-mode scenario (i.e. similar to the ITER reference scenario) has been the strongest affected and the benefit of high shaping to give good normalized confinement of H-98 similar to 1 at high Greenwald density fraction of f(GW) similar to 0.8 has not been recovered to date. In AUG with a full tungsten wall, a good normalized confinement H-98 similar to 1 could be achieved in the high triangularity baseline plasmas, albeit at elevated normalized pressure beta(N) > 2. The confinement lost with respect to the carbon devices can be largely recovered by the seeding of nitrogen in both JET and AUG. This suggests that the absence of carbon in JET and AUG with a metal wall may have affected the achievable confinement. Three mechanisms have been tested that could explain the effect of carbon or nitrogen (and the absence thereof) on the plasma confinement. First it has been seen in experiments and by means of nonlinear gyrokinetic simulations (with the GENE code), that nitrogen seeding does not significantly change the core temperature profile peaking and does not affect the critical ion temperature gradient. Secondly, the dilution of the edge ion density by the injection of nitrogen is not sufficient to explain the plasma temperature and pressure rise. For this latter mechanism to explain the confinement improvement with nitrogen seeding, strongly hollow Z(eff) profiles would be required which is not supported by experimental observations. The confinement improvement with nitrogen seeding cannot be explained with these two mechanisms. Thirdly, detailed pedestal structure analysis in JET high triangularity baseline plasmas have shown that the fuelling of either deuterium or nitrogen widens the pressure pedestal. However, in JET-ILW this only leads to a confinement benefit in the case of nitrogen seeding where, as the pedestal widens, the obtained pedestal pressure gradient is conserved. In the case of deuterium fuelling in JET-ILW the pressure gradient is strongly degraded in the fuelling scan leading to no net confinement gain due to the pedestal widening. The pedestal code EPED correctly predicts the pedestal pressure of the unseeded plasmas in JET-ILW within +/- 5%, however it does not capture the complex variation of pedestal width and gradient with fuelling and impurity seeding. Also it does not predict the observed increase of pedestal pressure by nitrogen seeding in JET-ILW. Ideal peeling ballooning MHD stability analysis shows that the widening of the pedestal leads to a down shift of the marginal stability boundary by only 10-20%. However, the variations in the pressure gradient observed in the JET-ILW fuelling experiment is much larger and spans a factor of more than two. As a result the experimental points move from deeply unstable to deeply stable on the stability diagram in a deuterium fuelling scan. In AUG-W nitrogen seeded plasmas, a widening of the pedestal has also been observed, consistent with the JET observations. The absence of carbon can thus affect the pedestal structure, and mainly the achieved pedestal gradient, which can be recovered by seeding nitrogen. The underlying physics mechanism is still under investigation and requires further understanding of the role of impurities on the pedestal stability and pedestal structure formation.
    Plasma Physics and Controlled Fusion 12/2013; 55(12):124043. DOI:10.1088/0741-3335/55/12/124043 · 2.39 Impact Factor
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    ABSTRACT: The enhancements carried out in the tokamak ASDEX Upgrade (AUG) are oriented toward the preparation of the future physics-related activities of ITER and DEMO. To address the main ITER issues, plasma configurations with a wider operational limit (e.g. higher triangularity) are planned for the future experimental campaigns in AUG. To evaluate the mechanical impact on the toroidal field (TF) magnet system a combined electromagnetic and structural finite element model was set up. At first extensive benchmarks of the models are carried out against the AUG reference design configurations with respect to stress [1-3], lateral displacement measurements and poloidal flux pattern. The numerical model was then applied to a set of actual high triangularity (HT) configurations generated by a more favorable poloidal field (PF) current distribution made possible by an extension of the power supply system. The resulting change of the poloidal flux pattern and the lateral force distribution has consequences for the coil shear stress and vault stability. Both aspects are monitored by a safety system measuring the PF flux placed on top and bottom of the outer surface of two TF coils (TFCs) between vault and the TFC supporting structure, so called Turn Over Structure (TOS). The range of the new HT configurations has induced a modification of the flux pattern, so that an adaptation of safety system is required to protect the TFCs system. Following the same criteria of the old safety system [4,5], a new set up of virtual coils will be integrated in the control system of AUG. The motivation of this new set up will be discussed in this paper.
    Fusion Engineering and Design 10/2013; 88(9-10):1541-1545. DOI:10.1016/j.fusengdes.2013.02.022 · 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
  • Nuclear Fusion 09/2013; 53(9):093018. DOI:10.1088/0029-5515/53/9/093018 · 3.24 Impact Factor
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    ABSTRACT: Amulti-machine database for the H-mode scrape-off layer power fall-off length, lambda(q) in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under the auspices of the International Tokamak Physics Activity. Regression inside the database finds that the most important scaling parameter is the poloidal magnetic field (or equivalently the plasma current), with lambda(q) decreasing linearly with increasing B-pol. For the conventional aspect ratio tokamaks, the regression finds lambda(q) alpha B-tor(-0.8). q(95)(1.1).P-SOL(0.1).R-geo(0), yielding lambda(q,) (ITER) congruent to 1mm for the baseline inductive H-mode burning plasma scenario at I-p = 15 MA. The experimental divertor target heat flux profile data, from which lambda(q) is derived, also yield a divertor power spreading factor (S) which, together with lambda(q), allows an integral power decay length on the target to be estimated. There are no differences in the lambda(q) scaling obtained from all-metal or carbon dominated machines and the inclusion of spherical tokamaks has no significant influence on the regression parameters. Comparison of the measured lambda(q) with the values expected from a recently published heuristic drift based model shows satisfactory agreement for all tokamaks.
    Nuclear Fusion 09/2013; 53(9):093031. DOI:10.1088/0029-5515/53/9/093031 · 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. DOI:10.1016/j.jnucmat.2013.01.011 · 2.02 Impact Factor
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    ABSTRACT: We report the latest results of turbulence and transport studies in the ASDEX Upgrade scrape-off layer (SOL). Dissimilarity between the plasma and the floating potential fluctuations is studied experimentally and by gyrofluid simulations. Measurements by a retarding field analyser reveal that both, edge-localized mode (ELM) and turbulent filaments, convey hot ions over large radial distances in the SOL. The measured far SOL ELM ion temperature increases with the ELM energy, consistent with earlier observations that large ELMs deposit a large fraction of their energy outside the divertor. In the SOL, the ELM suppression by magnetic perturbations (MPs) results in lower ELM ion energy in the far SOL. At the same time, large filaments of ion saturation current are replaced by more continuous bursts. Splitting of the divertor strike zones observed by the infrared imaging in H-mode with MPs agree with predictions from the EMC3-Eirene simulations. This suggests that the 'lobe' structures due to perturbation fields observed near the X-point are not significantly affected by plasma screening, and can be described by a vacuum approach, as in the EMC3-Eirene. Finally, some effects of the MPs on the L-mode SOL are addressed.
    Nuclear Fusion 07/2013; 53(7):073047. DOI:10.1088/0029-5515/53/7/073047 · 3.24 Impact Factor
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    ABSTRACT: In ASDEX Upgrade H-modes with H-98 approximate to 0.95, similar effect of the ICRF antenna loading improvement by local gas injection was observed as previously in L-modes. The antenna loading resistance R-a between and during ELMs can increase by more than 25% after a switch-over from a deuterium rate of 7.5.10(21) D/s injected from a toroidally remote location to the same amount of deuterium injected close to an antenna. However, in contrast to L-mode, this effect is small in H-mode when the valve downstream w.r.t. parallel plasma flows is used. In L-mode, a non-linearity of R-a at P-ICRF<30 kW is observed when using the gas valve integrated in antenna. Application of magnetic perturbations (MPs) in H-mode discharges leads to an increase of R-a>30% with no effect of spectrum and phase of MPs on R-a found so far. In the case ELMs are fully mitigated, the antenna loading is higher and steadier. In the case ELMs are not fully mitigated, the value of R-a between ELMs is increased. Looking at the W source modification for the improved loading, the local gas injection is accompanied by decreased values of tungsten (W) influx Gamma(W) from the limiters and its effective sputtering yield Y-W, with the exception of the locations directly at the antenna gas valve. Application of MPs leads to increase of Gamma(W) and Y-W for some of the MP phases. With nitrogen seeding in the divertor, ICRF is routinely used to avoid impurity accumulation and that despite enhanced Gamma(W) and Y-W at the antenna limiters.
    AIP Conference Proceedings 06/2013; 1580:271. DOI:10.1063/1.4864540
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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; DOI:10.1016/j.jnucmat.2013.01.005 · 2.02 Impact Factor

Publication Stats

4k Citations
472.00 Total Impact Points


  • 1993–2014
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 2001–2011
    • Rechenzentrum Garching (RZG) of the Max Planck Society and the IPP
      Arching, Bavaria, Germany
  • 2003
    • University of Helsinki
      Helsinki, Uusimaa, Finland
    • University of Toronto
      • Institute for Aerospace Studies
      Toronto, Ontario, Canada
    • Oak Ridge National Laboratory
      • Fusion Energy Division
      Oak Ridge, Florida, United States
  • 1995
    • Universität Stuttgart
      Stuttgart, Baden-Württemberg, Germany