A. Kallenbach

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

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Publications (319)519.2 Total impact

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    ABSTRACT: The far scrape-off layer transport is studied in ASDEX Upgrade H-mode discharges with high divertor neutral density , high power across the separatrix and nitrogen seeding to control the divertor temperature. Such conditions are expected for ITER but usually not investigated in terms of turbulent SOL transport. At high and the H-mode discharges enter a regime of high cross-field particle and power transport in the SOL which is accompanied by a significant change of the turbulence characteristic analogous to the transition from conductive to convective transport in L-mode. Parallel particle and power flux densities of several m−2 s−1 and 10 MW m−2 have been detected about ∼40 to 45 mm outside the separatrix mapped to the outer mid-plane. The particle flux fall-off length reached up to 45 mm. This paper presents for the first time an empirical condition to enter the high transport regime in H-mode and the relation of this regime to changes in the filamentary transport.
    Journal of Nuclear Materials 08/2015; 463. DOI:10.1016/j.jnucmat.2015.01.012 · 2.02 Impact Factor
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    ABSTRACT: ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density.
    Nuclear Fusion 06/2015; 55(6):063015. DOI:10.1088/0029-5515/55/6/063015 · 3.24 Impact Factor
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    ABSTRACT: With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 907–12), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 075020) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L–H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.
    Nuclear Fusion 06/2015; 55(6):063017. DOI:10.1088/0029-5515/55/6/063017 · 3.24 Impact Factor
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    ABSTRACT: In preparation of ITER operation, large machines have replaced their wall and divertor material to W (ASDEX Upgrade) or a combination of Be for the wall and W for the divertor (JET). Operation in these machines has shown that the influx of W can have a significant impact on the discharge evolution, which has made modelling of this impact for ITER an urgent task. This paper reports on such modelling efforts. Maximum tolerable W concentrations have been determined for various scenarios, both for the current ramp-up and flat-top phase. Results of two independent methods are presented, based on the codes ZIMPUR plus ASTRA and CRONOS, respectively. Both methods have been tested and benchmarked against ITER-like Ip RU experiments at JET. It is found that W significantly disturbs the discharge evolution when the W concentration approaches ~10−4; this critical level varies somewhat between scenarios.
    Nuclear Fusion 06/2015; 55(6):063031. DOI:10.1088/0029-5515/55/6/063031 · 3.24 Impact Factor
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    ABSTRACT: Detachment of high power discharges is obtained in ASDEX Upgrade by simultaneous feedback control of core radiation and divertor radiation or thermoelectric currents by the injection of radiating impurities. So far 2/3 of the ITER normalized heat flux Psep/R = 15 MW m−1 has been obtained in ASDEX Upgrade under partially detached conditions with a peak target heat flux well below 10 MW m−2. When the detachment is further pronounced towards lower peak heat flux at the target, substantial changes in edge localized mode (ELM) behaviour, density and radiation distribution occur. The time-averaged peak heat flux at both divertor targets can be reduced below 2 MW m−2, which offers an attractive DEMO divertor scenario with potential for simpler and cheaper technical solutions. Generally, pronounced detachment leads to a pedestal and core density rise by about 20-40%, moderate (<20%) confinement degradation and a reduction of ELM size. For AUG conditions, some operational challenges occur, like the density cut-off limit for X-2 electron cyclotron resonance heating, which is used for central tungsten control.
    Nuclear Fusion 05/2015; 55(5):053026. DOI:10.1088/0029-5515/55/5/053026 · 3.24 Impact Factor
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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
  • I. Zammuto · L. Giannone · A. Houben · A. Herrmann · A. Kallenbach
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    ABSTRACT: A long term project is started at the ASDEX Upgrade (AUG) tokamak aimed at the exploration of the compatibility of reduced activation ferritic/martensitic steel (RAFM) with fusion devices. The topic is oriented toward the preparation of future experiments such as ITER with its test blanket modules and DEMO with its first wall designed with RAFM. The goal of the project is to gather experience with ferromagnetic materials inside the vacuum vessel, dealing with magnetic perturbations, both in plasma and magnetic probes, and facing up the additional magnetic forces acting on the supporting structures. The project envisages a stepwise replacement of the traditional graphite tiles with ferritic steel. For the time being, the main AUG actor is the inner heat shield (IHS), but further development can be imagined in the future. Since 2013, two of the 15 tile rows of the IHS have been replaced with ferritic steel and since now the experimental campaign has not suffered any particular problem related to the perturbation field induced by the steel tiles, as predicted by the calculation.
    Fusion Engineering and Design 03/2015; DOI:10.1016/j.fusengdes.2015.01.048 · 1.15 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
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    ABSTRACT: The sources of neutrals at the outer midplane of the plasma are discussed. We find that both the flux of neutrals escaping the divertor through leaks and ion recycling at main chamber surfaces appear to contribute. The ion flux to the walls is larger than the flux entering the divertor and comparable to recycling at the divertor plate. The cause of these high wall ion fluxes is an enhancement of cross-field particle transport which gives rise to substantial convective heat transport at higher densities. We have further explored main chamber recycling and impurity transport utilizing a novel divertor 'bypass', which connects the outer divertor plenum to the main chamber. We find that leakage of neutrals (fuel and recycling impurities) from the divertor appears to be determined primarily by the conductance through the divertor structure, thus indicating that tight baffling would be desirable in a reactor for fuel and helium ash compression.
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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
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    ABSTRACT: Power exhaust in future fusion devices is critical and operation with a detached divertor is foreseen for ITER and DEMO. The evolution of detachment in nitrogen seeded H-mode discharges at ASDEX Upgrade is categorized in four phases. Complete detachment of the outer target is found to be correlated with a strongly localized radiation at the X-point and a pressure loss at the pedestal top at almost constant core plasma pressure. SOLPS modeling shows that enhanced radial transport in the divertor region is necessary to reconcile the experimental profiles with the simulations. The modeling supports the experimental observation of the correlation of complete detachment with an X-point radiation and a reduction of the pedestal top pressure. A remaining discrepancy are significantly lower neutral densities in the divertor compared to experiment. The effects of wall pumping, the particle reflection model and the boundary conditions on the plasma solution are discussed.
    Journal of Nuclear Materials 12/2014; 463. 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: ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs in both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically, the Be PFCs of JET led to a very small low-Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady-state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by codeposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N-2) (AUG and JET) or by combining N-2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N-2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fueling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however, in both devices the W content is increased during ICRH operation most probably due to increased W sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20%-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not yet clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N-2 seeding observed in both devices hints to the fact, that low-Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.
    IEEE Transactions on Plasma Science 03/2014; 42(3):552-562. DOI:10.1109/TPS.2014.2298253 · 0.95 Impact Factor
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    ABSTRACT: A new set of diagnostics has been implemented on ASDEX Upgrade to measure the input impedance of the ICRF antennas, in the form of a voltage and current probe pair installed on each feeding line of every antenna. Besides allowing the measurement of the reflection coefficient Gamma of each antenna port, the probes have two advantages: first, they are located close to the antenna ports (similar to 3 m) and thus the measurements are not affected by the uncertainties due to the transmission and matching network; second, they are independent of matching conditions. These diagnostics have been used to study the behavior of the ASDEX Upgrade antennas while changing the plasma shape (low to high triangularity) and applying magnetic perturbations (MPs) via saddle coils. Scans in the separatrix position R-sep were also performed. Upper triangularity delta(o) was increased from 0.1 to 0.3 (with the lower triangularity delta(o) kept roughly constant at 0.45) and significant decreases in vertical bar Gamma vertical bar (up to similar to 30%, markedly improving antenna coupling) and moderate changes in phase (up to similar to 5 degrees) off on each feeding line were observed approximately at delta(o) >= 0.29. During MPs (in similar to 0.5 s pulses with a coil current of 1 kA), a smaller response was observed: 6% - 7% in vertical bar Gamma vertical bar, with changes in phase of 5 apparently due to R p scans only. As 1 is usually in the range 0.8 - 0.9, this still leads to a significant increase in possible coupled power. Numerical simulations of the antenna behavior were carried out using the FELICE code; the simulation results are in qualitative agreement with experimental measurements. The results presented here complement the studies on the influence of gas injection and MPs on the ICRF antenna performance presented in [4].
    20th Topical Conference on Radio Frequency Power in Plasmas; 02/2014
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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: For the design and operation of large fusion devices, a detailed understanding of the power exhaust processes is necessary. This paper will give an overview of the current research on divertor power load mechanisms. The results shown are obtained in JET with the ITER-like wall (ILW) and ASDEX-Upgrade with tungsten coated plasma-facing components (PFCs). The challenges of infrared thermography on an ITER-like bulk tungsten divertor are presented. For the steady-state heat load, the power fall-off length lambda(q) in JET-ILW is compared to an empirical scaling found in JET and the ASDEX-Upgrade with carbon PFCs. A first attempt to scale the divertor broadening S in the ASDEX-Upgrade with tungsten PFCs is shown. The edge localized mode (ELM) duration t(ELM) in JET-C and JET-ILW is compared. For similar pedestal conditions (T-e,T-ped and n(e,ped)), similar ELM durations are found in JET-C and JET-ILW. For higher n(e,ped) at the same pedestal pressure p(e,ped), longer ELM durations are found in JET-ILW. The pedestal pressure p(e,ped) is found to be a good qualifier for the ELM energy fluency in both JET-C and JET-ILW. Improved diagnostic capabilities reveal ELM substructures on the divertor target occurring a few milliseconds before the ELM crash.
    Plasma Physics and Controlled Fusion 12/2013; 55(12):124039. DOI:10.1088/0741-3335/55/12/124039 · 2.39 Impact Factor

Publication Stats

4k Citations
519.20 Total Impact Points

Institutions

  • 1993–2015
    • 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
    • Princeton University
      Princeton, New Jersey, United States
  • 1995
    • Universität Stuttgart
      Stuttgart, Baden-Württemberg, Germany