Impurity behaviour in the ASDEX Upgrade divertor tokamak with large area tungsten walls
TL;DR: In this paper, an area of 5.5 m2 of graphite tiles was replaced by tungsten-coated tiles representing about two-thirds of the total area of the central column.
Abstract: At the central column of ASDEX Upgrade, an area of 5.5 m2 of graphite tiles was replaced by tungsten-coated tiles representing about two-thirds of the total area of the central column. No negative influence on the plasma performance was found, except for internal transport barrier limiter discharges. The tungsten influx ΓW stayed below the detection limit only during direct plasma wall contact or for reduced clearance in divertor discharges spectroscopic evidence for ΓW could be found. From these observations a penetration factor of the order of 1% and effective sputtering yields of about 10-3 could be derived, pointing to a strong contribution by light intrinsic impurities to the total \mbox{W-sputtering}. The tungsten concentrations ranged from below 10-6 up to a few times 10-5. Generally, in discharges with increased density peaking, a tendency for increased central tungsten concentrations or even accumulation was observed. Central heating (mostly) by ECRH led to a strong reduction of the central impurity content, accompanied by a very benign reduction of the energy confinement. The observations suggest that the W-source strength plays only an inferior role for the central W-content compared to the transport, since in the discharges with increased W-concentration neither an increase in the W-influx nor a change in the edge parameters was observed. In contrast, there is strong experimental evidence, that the central impurity concentration can be controlled externally by central heating.
Citations
More filters
Massachusetts Institute of Technology1, European Atomic Energy Community2, University of Toronto3, Japan Atomic Energy Agency4, Max Planck Society5, Forschungszentrum Jülich6, Princeton Plasma Physics Laboratory7, University of Wisconsin-Madison8, University of California, San Diego9, Lawrence Livermore National Laboratory10, Institut national de la recherche scientifique11, École Polytechnique Fédérale de Lausanne12, Nagoya University13
TL;DR: In this paper, the authors describe the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER and compare their predictions with the new experimental results.
Abstract: Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.
943 citations
TL;DR: The physics model of electron cyclotron heating (ECH) and current drive (ECCD) is becoming well validated through systematic comparisons of theory and experiment, leading to applications including stabilization of magnetohydrodynamic instabilities like neoclassical tearing modes, control and sustainment of desired profiles of current density and plasma pressure, and studies of localized transport in laboratory plasmas as discussed by the authors.
Abstract: The physics model of electron cyclotron heating (ECH) and current drive (ECCD) is becoming well validated through systematic comparisons of theory and experiment. This work has shown that ECH and ECCD can be highly localized and robustly controlled in toroidal plasma confinement systems, leading to applications including stabilization of magnetohydrodynamic instabilities like neoclassical tearing modes, control and sustainment of desired profiles of current density and plasma pressure, and studies of localized transport in laboratory plasmas. The experimental work was supported by a broad base of theory based on first principles which is now well encapsulated in linear ray tracing codes describing wave propagation, absorption, and current drive and in fully relativistic quasilinear Fokker–Planck codes describing in detail the response of the electrons to the energy transferred from the wave. The subtle balance between wave-induced diffusion and Coulomb relaxation in velocity space provides an understandin...
276 citations
01 Oct 2007
TL;DR: In this paper, the authors describe the status of investigations on the use of tungsten as a first wall material in a fusion reactor. But, due to the high erosion rate of tritium, carbon plasma facing components (PFCs) appear to be unacceptable for a commercial fusion reactor, therefore, they are not suitable for the first wall coverage.
Abstract: The observation in JET of co-deposition of tritium with carbon has led to a broad discussion on the replacement of graphite by a high-Z material for the first wall coverage. Moreover, due to the high erosion rate, carbon plasma facing components (PFCs) appear to be unacceptable for a commercial fusion reactor. Research in this area has subsequently gained increased attention. This paper describes the status of investigations on the use of tungsten as a first wall material. It discusses on the physical side the plasma wall interaction, the transport of tungsten in the plasma boundary and in the core. As an intermediate step on the technological side, graphite is often coated with tungsten layers. For highly loaded surfaces in a fusion reactor finally bulk tungsten components will have to be developed.
141 citations
TL;DR: The sawtooth instability in tokamak plasmas results in a periodic reorganization of the core plasma as discussed by the authors, which can lead to the early triggering of neo-classical tearing modes (NTMs) at low plasma pressure.
Abstract: The sawtooth instability in tokamak plasmas results in a periodic reorganization of the core plasma. A typical sawtooth cycle consists of a quiescent period, during which the plasma density and temperature increase, followed by the growth of a helical magnetic perturbation, which in turn is followed by a rapid collapse of the central pressure. The stabilizing effects of fusion-born ? particles are likely to lead to long sawtooth periods in burning plasmas. However, sawteeth with long quiescent periods have been observed to result in the early triggering of neo-classical tearing modes (NTMs) at low plasma pressure, which can, in turn, significantly degrade confinement. Consequently, recent experiments have identified various methods to deliberately control sawtooth oscillations in an attempt to avoid seeding NTMs whilst retaining the benefits of small, frequent sawteeth, such as the prevention of core impurity accumulation. Sawtooth control actuators include current drive schemes, such as electron cyclotron current drive, and tailoring the fast ion population in the plasma using neutral beam injection or ion cyclotron resonance heating.
116 citations
TL;DR: In this article, the effects of poloidal asymmetries and heated minority species are shown to be necessary to accurately describe heavy impurity transport in present experiments in JET and ASDEX Upgrade.
Abstract: The effects of poloidal asymmetries and heated minority species are shown to be necessary to accurately describe heavy impurity transport in present experiments in JET and ASDEX Upgrade. Plasma rotation, or any small background electrostatic field in the plasma, such as that generated by anisotropic external heating can generate strong poloidal density variation of heavy impurities. These asymmetries have recently been added to numerical tools describing both neoclassical and turbulent transport and can increase neoclassical tungsten transport by an order of magnitude. Modelling predictions of the steady-state two-dimensional tungsten impurity distribution are compared with tomography from soft x-ray diagnostics. The modelling identifies neoclassical transport enhanced by poloidal asymmetries as the dominant mechanism responsible for tungsten accumulation in the central core of the plasma. Depending on the bulk plasma profiles, turbulent diffusion and neoclassical temperature screening can prevent accumulation. Externally heated minority species can significantly enhance temperature screening in ICRH plasmas.
115 citations
References
More filters
TL;DR: In this paper, two main regimes of operation are foreseen namely ELMy H-mode and steady-state operation, the latter in discharges with internal transport barriers (ITB).
257 citations
TL;DR: In this article, it was shown that deposition and H-isotope retention in JET is highly asymmetric, with deposition predominantly in the inner divertor, where flaking deposits form on water-cooled louvres shadowed from the plasma.
200 citations
TL;DR: In this paper, the dependence of core plasma impurity transport on the Z number has been investigated for ASDEX upgrade H?mode discharges and the calculated diffusion coefficient and drift velocity are close to the experimental values for lower-Z elements Ne and Ar.
Abstract: The dependence of core plasma impurity transport on the Z number has been investigated for ASDEX Upgrade H?mode discharges. For the elements Ne, Ar, Kr and Xe the diffusion coefficient in the centre is D ? 6 ? 10-2m2/s and rises with the radial distance from the centre. With increasing Z number the transport becomes strongly convective with inward directed drift velocities that produce very peaked impurity densities for high Z. The inward drift decreases with decreasing deuterium density gradient. Neoclassical transport of the impurities has been calculated numerically. The calculated diffusion coefficient and drift velocity are close to the experimental values for the lower-Z elements Ne and Ar. However, for high Z, the calculated diffusion coefficient is smaller by a factor of up to 2.5 and the inward drift velocity is too small by a factor of 10. Toroidal rotation of the plasma that leads to an increased impurity density on the outboard side of the flux surfaces is not taken into account by the neoclassical calculations. Inboard/outboard asymmetries are not present for Ar and Ne with toroidal Mach number Mtor around 1. However, for heavier elements than Kr with Mtor ? 2 and an outboard/inboard ratio of ? 1.5, poloidal variation of the impurity density is important and might account for the discrepancy between the measured and calculated transport coefficients.
128 citations
TL;DR: In this paper, Janeschitz et al. analyzed the effects of tritium codeposition on the first-wall and divertor of a reduced-size ITER-FEAT with a strike-point carbon divertor target and metallic walls.
124 citations
TL;DR: Tungsten divertor plates have been used in ASDEX Upgrade for a full experimental campaign of approximately 800 discharges as mentioned in this paper, and the results obtained so far are presented and the implications with respect to the construction of future fusion devices are discussed.
123 citations