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ABSTRACT: Simple models for radial transport in the bulk plasma, for hydrogen recycling in a poloidal divertor and for the scrape-off layer as a link between these two regions are combined and used to study hydrogen refuelling and helium pumping in a thermonuclear tokamak reactor. It is shown that the helium pumping requirements impose restrictions on the hydrogen refuelling depth and the minimum hydrogen flux to be pumped together with helium. Pellet refuelling with penetration of only a fraction of the minor radius seems to be adequate for INTOR.
Nuclear Fusion 01/2011; 25(1):89. · 4.09 Impact Factor
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K. Steinmetz,
H. Niedermeyer,
J.-M. Noterdaeme,
F. Wagner,
F. Wesner,
J. Bäumler,
G. Becker,
W. Becker,
H.S. Bosch,
M. Brambilla, [......],
G. Siller,
P. Smeulders,
M. Söll,
E. Speth,
F.X. Söldner,
A. Stäbler,
K.-H. Steuer,
O. Vollmer,
H. Wedler,
D. Zasche
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ABSTRACT: The paper summarizes the experiments performed with ion cyclotron resonance heating (ICRH) on ASDEX, from November 1984 until March 1986; the most interesting results are reported and discussed in detail. Heating and confinement studies using the hydrogen second harmonic scheme and the hydrogen minority scheme (PIC < 2.6 MW, tIC < 1.5 s) show a typical L-mode behaviour, i.e. a power dependent confinement degradation, which is rather similar to that found with neutral beam injection (NBI) heating. ICRH is accompanied by a slightly improved particle and energy confinement compared with that of NBI; this is also true for a combined ICRH + NBI scheme, up to Ptot ≈ 4.5 MW, absorbed in the plasma. Particular efforts have been devoted to investigations of the second harmonic regime in H/D plasmas with nH/ne ≈ 0.1 - 1, with a view to heating mixtures in reactor relevant plasmas. The achievement of H-mode transitions with ICRH alone in the hydrogen minority scheme at an absorbed RF power of about 1.1 MW supports the assumption of common confinement properties in auxiliary heated tokamaks, since they appear to be widely independent of the additional heating method. ICRH specific impurity problems, such as the strong release of iron from the vessel walls, have been overcome by applying extensive in situ wall carbonization. The mechanisms responsible for impurity generation have partly been identified and analysed; however, the problem still remains to be solved. Impurities preferentially released from the ICRH antenna do not pose problems.
Nuclear Fusion 01/2011; 29(2):277. · 4.09 Impact Factor
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M. Bessenrodt-Weberpals,
K. McCormick,
F.X. Söldner,
F. Wagner,
H.S. Bosch,
O. Gehre,
E.R. Müller,
H.D. Murmann, J. Neuhauser,
W. Poschenrieder,
K.-H. Steuer,
N. Tsois,
ASDEX Team
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ABSTRACT: The scaling of the energy confinement time with plasma density and current has been investigated for ohmically heated tokamak discharges in ASDEX. The linear dependence τE e is maintained in the high density improved Ohmic confinement (IOC) regime with peaked density profiles. The peaking of the radial density profile can be brought about by reducing the net power flow through the plasma surface, thereby leading to a reduction of the edge density. Tailoring of the radiation profile with the addition of low-Z impurities, for example neon, gives access to the IOC regime under conditions where otherwise the degraded saturated Ohmic confinement (SOC) behaviour prevails. The energy confinement time τE increases with current and decreases with heating power also in Ohmic discharges, as is shown by a statistical analysis. However, with the intrinsic coupling between power and current, the two relationships cancel and τE becomes independent of POH and Ip. The two most prominent features of Ohmic confinement can therefore be explained on the basis of simple physical models.
Nuclear Fusion 01/2011; 31(1):155. · 4.09 Impact Factor
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ABSTRACT: The electron thermal conductivity χe is calculated from the evolution of the heat pulse immediately after a sawtooth crash as a forced boundary problem. The electron cyclotron emission signal, measured just outside the sawtooth mixing radius, is used as the time dependent boundary condition. This method makes no assumptions about the plasma behaviour inside the mixing radius and permits the accurate determination of the radial dependence of χe.
Nuclear Fusion 01/2011; 28(9):1509. · 4.09 Impact Factor
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ABSTRACT: The thermal stability of a collisional, radiating edge plasma in magnetic confinement experiments with respect to marfe-type perturbations is investigated analytically and numerically. The fluid description used includes a non-local electron heat flux model and non-equilibrium impurity radiation cooling. The dispersion relation of a homogeneous plasma is derived analytically, and stability criteria are given. Various experimental scenarios are simulated with a sophisticated numerical model including all the relevant dynamics along field lines and a rough approximation of cross-field transport. The formation and history of global 'marfes' and the onset of relaxation oscillations under appropriate conditions are demonstrated. The possible occurrence of more localized perturbations is discussed. The results compare well with those of experiments.
Nuclear Fusion 01/2011; 26(12):1679. · 4.09 Impact Factor
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M. Kaufmann,
K. Büchl,
G. Fussmann,
O. Gehre,
K. Grassie,
O. Gruber,
G. Haas,
G. Janeschitz,
M. Kornherr,
K. Lackner, [......],
K. McCormick,
V. Mertens, J. Neuhauser,
H. Niedermeyer,
W. Sandmann,
W. Schneider,
D. Zasche,
H.-P. Zehrfeld,
Z.A. Pietrzyk,
G. Vlases
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ABSTRACT: Tokamak discharges with repetitive pellet fuelling were investigated in the ASDEX divertor device. The importance of sufficiently high divertor recycling for a high density at the separatrix and for successful density buildup in the bulk plasma was demonstrated. In contrast to low recycling discharges where no permanent improvement of the energy confinement time was achieved, in OH-heated discharges with high recycling an energy confinement time of 160 ms was reached, the normal value being 80 ms in the rollover region. The peaked density profiles in this case were accompanied by reduced or suppressed sawtooth activity and finally ended in a phase of strong central impurity accumulation. The particle transport was characterized by strong, non-classical inward drift, while the improved energy transport can be explained by the following alternatives: (la) a local model which assumes neo-Alcator χe for the electrons and χi= 3χneocl for the ions in the gas puff cases, reducing to χi= χneocl for the optimum pellet cases; (1b) the assumption χi= χneocl under all conditions and an electron energy confinement worse than neo-Alcator in the rollover region in gas-puff-discharges; (2) a profile consistency picture where Te(a) determines the energy confinement. Low power, NI heated discharges with pellet fuelling behave like Ohmic discharges, while for high power in the L-mode no successful density buildup was possible, and τE was not improved. The H-regime was extended from a density maximum e = 0.8 × 1020 m-3 without pellets to e = 1.2 × 1020 m-3 by the injection of pellets. In this case a density buildup takes place, but further density profile peaking could not be observed.
Nuclear Fusion 01/2011; 28(5):827. · 4.09 Impact Factor
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ABSTRACT: The m = 2 mode stability is investigated in the ISAR T1-B toroidal high-beta stellarator (major diameter 2.7 m, capacitor bank energy 0.5 or 1.5 MJ) for a great variety of plasma parameters using deuterium filling gas (0.5 < ωL τii < 200; 0.1 < rL/a < 0.35; ωL = ion Larmor frequency; rL = ion Larmor radius; a = plasma radius; τii = ion self-collision time). The ion mass dependence was checked by a small number of discharges with hydrogen. For rL/a < ηcrit, ηcrit = 0.14, fast-growing m = 2 modes were found, the limit ηcrit being essentially independent of the collision frequency within the accessible range (2 ωL τii 20). For rL/a > 0.14 no fast-growing m = 2 modes were observed, the plasma lifetime being limited by a m = 1, k ≈ 0 mode. The normalized growth rate, = γ/(Vi/a) = 0.15 ± 0.05, (Vi = ion thermal speed) was about half the sharp boundary value and slightly higher than that calculated for a diffuse pressure profile. The experimentally determined minimum value of rL/a, for which the m = 2 mode is stabilized, is by about a factor of three smaller than Freidberg's estimation using a sharp-boundary Vlasov model. Advanced calculations in a sharp boundary model by Lewis and Turner deviate in the critical limit only by a factor of about 1.5 and recent results using a diffuse pressure profile by Herrnegger and Schneider and by Freidberg and Hewett agree quite well with the experimental data.
Nuclear Fusion 01/2011; 17(1):3. · 4.09 Impact Factor
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ABSTRACT: The radial transport of oxygen at the edge of a tokamak plasma is considered. The resonant charge exchange between oxygen and hydrogen (O+ + H O + H+) as well as the ionization and disintegration of oxygen molecules are taken into account. Approximate analytic expressions for these effects as obtained from a kinetic model are added to the one-dimensional oxygen rate equations and solved numerically using a fast algorithm developed recently. A significantly enhanced penetration of oxygen into the central plasma as well as an increased oxygen recycling at the edge are obtained, in particular in the vicinity of a limiter, with high local hydrogen and oxygen fluxes. The non-resonant charge exchange of higher oxygen charge states with hydrogen neutrals changes their distribution and radiation in the central plasma but has little effect on the oxygen transport at the periphery. A simple analytical model has been developed to describe oxygen impurity transport in the scrape-off layer and the results are compared with numerical simulations.
Nuclear Fusion 01/2011; 24(8):989. · 4.09 Impact Factor
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ABSTRACT: The wavenumber and frequency spectrum of LHD-instabilities were measured by coherent CO2 laser light scattering in the boundary layer of a theta-pinch plasma. The 9-m-long theta pinch produced a stationary plasma for up to 20 μs. These conditions allowed the LHD-wave to be examined in the low-drift-velocity regime and a comparison to be carried out with the linearized theory. The fluctuation density amplitudes and the wavenumber spectrum as functions of the discharge parameters indicated that ion trapping is likely to be the limiting saturation mechanism.
Nuclear Fusion 01/2011; 21(2):257. · 4.09 Impact Factor
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ABSTRACT: The impurity flow parallel to the magnetic field lines in a collisional tokamak scrape-off layer is numerically investigated. The rate equations are solved treating each ionization state as a test fluid which interacts with the given hydrogen background plasma via collisions and the ambipolar electric field. Results for typical impurities (O, Fe, He, etc.) show that collisional friction usually forces the impurities to flow nearly at hydrogen speed. Thermal forces, however, can become dominant locally for small Mach number and large temperature gradient (e.g. strong target recycling), causing impurity flow reversal and subsequent accumulation outside the recycling region. The criterion for flow reversal is roughly M <λi/λT where λi is the hydrogen ion mean free path and λT is the temperature gradient length. Self-sputtering at the target plates is calculated, showing the importance of fractional impurity acceleration in addition to the charge-statedependent electrostatic energy gain.
Nuclear Fusion 01/2011; 24(1):39. · 4.09 Impact Factor
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ABSTRACT: Experiments with the first complete high-β stellarator torus with helical magnetic axis are reported. The Garching 2.6-MJ capacitor bank was used to feed a toroidal theta coil with a major.diameter of 2.7 m (ISAR T1), Toroidal equilibrium was achieved by superposing helical ℓ = 1 and ℓ = 2 (partly also ℓ = 0) fields on the slender (A = RT/r0 ≈ 150) toroidal pinch plasma such that the plasma surface facing the torus centre was more corrugated than the outer side (M-and-S effect). The toroidal equilibrium condition and the corresponding plasma distortions were consistent with sharp-boundary-model predictions. Effects in connection with initial dynamics, toroidal plasma currents and transverse magnetic fields could be explained by simple models. In agreement with sharp-boundary theory, short wavelength m = 1 and m = 2 modes were found to be stable and long wavelength m = 1 modes were unstable, limiting the plasma life-time by wall contact. Long-wavelength m ≥ 2 instabilities were not observed in contrast to sharp-boundary theory, i.e. this model is much too pessimistic for m ≥ 2 modes, even if the finite gyroradius is included. No significant difference in the stability behaviour was found, compared with previous linear and toroidal sector experiments.
Nuclear Fusion 01/2011; 15(1):133. · 4.09 Impact Factor
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J. Adamek,
C. Angioni,
G. Antar,
C. V. Atanasiu,
M. Balden,
W Becker,
K. Behler,
K. Behringer,
A Bergmann,
R. Bilato, [......],
C. Wigger,
M. Wischmeier,
E. Wolfrum,
E. Würsching,
D. Yadikin,
Q Yu,
D. Zasche,
T. Zehetbauer,
M. Zilker,
H. Zohm
The Review of scientific instruments 03/2010; 81(3):039903. · 1.52 Impact Factor
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J Adamek,
C Angioni,
G Antar,
C V Atanasiu,
M Balden,
W Becker,
K Behler,
K Behringer,
A Bergmann,
R Bilato, [......],
C Wigger,
M Wischmeier,
E Wolfrum,
E Wuersching,
D Yadikin,
Q Yu,
D Zasche,
T Zehetbauer,
M Zilker,
H Zohm
Reviews of Scientific Instruments. 01/2010; 81(3):-.
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ABSTRACT: Refuelling of fusion plasmas by the injection of frozen hydrogen isotope pellets from the magnetic high field side (HFS) rather than from the low field side (LFS) increases the fuelling efficiency due to the grad B drift accelerating the ablatant material in the positive major radius direction, thus towards the centre of the plasma. The HFS pellet fuelling is therefore presently established as the main fuelling scenario for ITER. This paper describes an HFS pellet database developed at the ASDEX Upgrade tokamak and the HFS penetration depth scaling derived by statistical analysis performed on the dataset. Also a comparison is made with the existing empirical LFS penetration depth scaling and theoretical models.
Nuclear Fusion 05/2008; 48(6):065009. · 4.09 Impact Factor
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J. Neuhauser,
V. Bobkov,
G.D. Conway,
R. Dux,
T. Eich,
M. Garcia-Munoz,
A. Herrmann,
L.D. Horton,
A. Kallenbach,
S. Kalvin, [......],
H.D. Murmann,
R. Neu,
A.G. Peeters,
M. Reich,
V. Rohde,
A. Schmid,
W. Suttrop,
M. Tsalas,
E. Wolfrum,
the ASDEX Upgrade Team
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ABSTRACT: In order to assess the contribution of edge localized modes (ELMs) to plasma–wall interaction in future fusion experiments like ITER, a sound experimental database for model validation and extrapolation, and, to be prepared for the unfavourable case, the development of tools for ELM mitigation are required. On ASDEX Upgrade a large amount of experimental information has been accumulated from various diagnostics on the structure and dynamics of natural as well as pellet induced ELMs, and on related wall effects. In this paper a survey of type-I ELM results is given first and recent progress is then described in detail. In between ELMs, strong mode activity is observed in a wide mode number and frequency range, specifically large amplitude (~20%) low frequency (several kilohertz) fluctuations. The initial dynamic ELM phase is dominated by the rapid growth of helical, low mode number structures rotating in the pedestal E × B direction, while the subsequent saturation and profile erosion phase is more complex and scenario dependent. Bursts of filaments ejected from the hot edge into the scrape-off layer are correlated with primary pedestal mode rotation. After partial edge profile collapse, a quiescent recovery phase is obtained despite substantial residual edge gradients. Pellet induced ELMs behave similarly to spontaneous ones, at least for the smallest pellets available so far.
Nuclear Fusion 02/2008; 48(4):045005. · 4.09 Impact Factor
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R Neu,
M Balden,
V Bobkov,
R Dux,
O Gruber,
A Herrmann,
A Kallenbach,
M Kaufmann,
C F Maggi,
H Maier, [......],
D Wagner,
M Wischmeier,
E Wolfrum,
E Würsching,
D Yadikin,
Q Yu,
D Zasche,
T Zehetbauer,
M Zilker,
H Zohm
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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.
Plasma Physics and Controlled Fusion 11/2007; 49(12B):B59. · 2.42 Impact Factor
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ABSTRACT: Pellets injected into type-I ELMy H-mode discharges are known to trigger edge-localized modes (ELMs). In order to understand the underlying processes the triggering mechanism was investigated in this paper. The major questions of the investigations to be answered were: at which magnetic surface was the ELM initiated and what was the corresponding perturbation caused by the ablating pellet? During the investigations the natural ELM cycle was probed by injecting pellets from the high field side of the ASDEX Upgrade tokamak with significantly lower frequency than the natural ELM frequency. To determine the location of the seed perturbation of the ablating pellet triggering an ELM, the dynamics of the triggered ELMs was linked to the time history of the pellet position in the plasma. The ELM onset was determined by analysing magnetic pick-up coil signals and its delay relative to the time when the pellet crossed the separatrix was measured as a function of the pellet velocity. Supposing that to trigger an ELM a pellet has to reach a certain magnetic surface of the plasma independently of its mass and velocity, the most probable location of the seed perturbation was found to be at the middle of the pedestal—at the high plasma pressure gradient region. The onset of the MHD signature of the ELMs was detected about 50 µs after the pellet reached the seed position. According to our observations ELMs can be triggered either by the cooling of the pedestal region causing sudden increase of the pedestal plasma pressure gradient driving the plasma to the unstable region of the ballooning mode or by the strong MHD perturbation triggering an instability developing into an ELM.
Nuclear Fusion 08/2007; 47(9):1166. · 4.09 Impact Factor
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ABSTRACT: The pellet pacing ELM control concept developed at the mid-sized tokamak ASDEX Upgrade is considered for the larger machines JET and ITER as well. By driving up the ELM frequency, the ELM size can be reduced and eventually suppressed below a dangerous level. An according pellet injection system for JET is under construction; the ITER design includes one as well. However, it is still to be proven whether the concept will also work for bigger machine sizes. A step forward in this mission was achieved by re-analysing previous JET experiments dedicated to pellet particle fuelling. Although the experiments were performed in a parameter regime unsuitable for pellet ELM pacing both with respect to pellet frequency and size they demonstrate that prompt triggering of ELMs takes place at JET as well. Prompt triggering, an indispensable premise for useful pacing, obviously can be achieved already with a pellet size sufficiently small to avoid unbearable side-effects by particle fuelling. An ELM released by the local perturbation imposed by the ablating pellet in the plasma edge region is detected when only a minor fraction (less than 1%) of the fuelling size pellets (containing about 4 × 1021 D) was ablated and deposited. It is thus very likely that JET will allow for investigations on the operational technique and the underlying physics of pellet pacing once the new system becomes operational.
Nuclear Fusion 07/2007; 47(8):754. · 4.09 Impact Factor
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A.V. Chankin,
D.P. Coster,
N. Asakura,
X. Bonnin,
G.D. Conway,
G. Corrigan,
S.K. Erents,
W. Fundamenski,
J. Horacek,
A. Kallenbach,
M. Kaufmann,
C. Konz,
K. Lackner,
H.W. Müller, J. Neuhauser,
R.A. Pitts,
M. Wischmeier
[show abstract]
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ABSTRACT: Examination of radial electric field (Er) profiles in the scrape-off layer (SOL) of ASDEX Upgrade (AUG) and JET revealed large discrepancies between 2D fluid edge modelling and experiment. Experimental profiles of plasma potential (Vp) in the outer (low field) side of the plasma, obtained with reciprocating Langmuir probes, decay radially with electron temperature, Te, with the −eEr/∇Te ratio being > 1.5. In contrast, code simulated Er are fairly low in most of the SOL (compared with −∇Te/e). Modelling with kinetic treatment of neutrals and drifts was performed using the SOLPS code for AUG cases and EDGE2D-Nimbus for JET cases.Mismatches between modelled and experimental Er may be caused by the recently established tendency for the SOLPS code to underestimate Te in the divertor of AUG. It was attributed to non-locality of parallel transport of supra-thermal, heat-carrying electrons originating upstream of the divertor, which are usually only weakly collisional and can penetrate, with few collisions, to the target. Ratios −eEr/∇Te obtained from the probe measurements in JET are of order 1.6, while in AUG, JT-60U and TCV they are of order 3. Such high values point to the possibility of fast electrons contributing, apart from target heat fluxes, also to the formation of the Debye sheath.The problem of the underestimation of Er in the codes must be closely related with the well-known problem of the underestimation of those parts of parallel ion flows in the SOL that are influenced by the toroidal field direction. It was demonstrated earlier that parallel ion flow at the outer midplane is dominated by the ion Pfirsch–Schlüter flow, which in turn is partly driven by the radial electric field. The Te and Er discrepancies, as well as discrepancies between simulated and experimental parallel ion flows, raise a question of the validity of fluid codes for the plasma edge modelling and prompt the inclusion of kinetic effects into present-day 2D fluid codes which assume strong collisionality.
Nuclear Fusion 05/2007; 47(5):479. · 4.09 Impact Factor
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ABSTRACT: A reciprocating probe in the lower divertor of ASDEX Upgrade, capable of accessing the low-field (LFS) and high-field side (HFS) scrape-off layers (SOLs) as well as the private flux region, was equipped with a Mach probe and used to measure flows in the vicinity of the lower x-point. We report on our measurements from ohmic and low-power H-mode discharges with ion B × ∇B drift towards the bottom x-point, and discuss their relevance to the current SOL/divertor flow understanding. In ohmic discharges, we present the evolution of divertor SOL and private flux flow profiles for increasing central ne. We show that the private flux flow is mainly directed from the HFS to the LFS at low densities. At medium-high densities the flow profile becomes more symmetric, and at very high densities the flow direction reverses on the LFS separatrix, having a LFS to HFS direction inside the private flux. We discuss the possible mechanisms that could affect divertor flows and produce such behaviour and conclude that pressure asymmetry between the two divertor legs combined with an E × B drift towards the inner divertor is a likely driving mechanism. At the HFS SOL, very large Mach numbers (typically exceeding M = 1) were observed in most cases. In low-power H-mode discharges inter-ELM flows were observed to be very similar to ohmic ones.
Plasma Physics and Controlled Fusion 05/2007; 49(6):857. · 2.42 Impact Factor
