Filamentation of tokamak plasmas

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Identification of the ubiquitous Coriolis momentum pinch in JET tokamak plasmas

A broad survey of the experimental database of neutral beam heated plasmas in the JET tokamak has established the theoretically expected ubiquity, in rotating plasmas, of a convective transport mechanism which has its origin in the vertical particle drift resulting from the Coriolis force. This inward convection, or pinch, leads to inward transport of toroidal angular momentum and is characterized by pinch numbers RV/¿, which rise from near unity at r/a ˜ 0.25 to around 5 at r/a ˜ 0.85. Linear gyrokinetic calculations of the Coriolis pinch number and the Prandtl number ¿/¿i are in good agreement with the experimental observations, with similar dependences on plasma parameters. The data, however, do not rule out contributions from different processes, such as residual stresses

Global performance enhancements via pedestal optimisation on ASDEX Upgrade

Results of experimental scans of heating power, plasma shape, and nitrogen content are presented, with a focus on global performance and pedestal alteration. In detailed scans at low triangularity, it is shown that the increase in stored energy due to nitrogen seeding stems from the pedestal. It is also shown that the confinement increase is driven through the temperature pedestal at the three heating power levels studied. In a triangularity scan, an orthogonal effect of shaping and seeding is observed, where increased plasma triangularity increases the pedestal density, while impurity seeding (carbon and nitrogen) increases the pedestal temperature in addition to this effect. Modelling of these effects was also undertaken, with interpretive and predictive models being employed. The interpretive analysis shows a general agreement of the experimental pedestals in separate power, shaping, and seeding scans with peeling-ballooning theory. Predictive analysis was used to isolate the individual effects, showing that the trends of additional heating power and increased triangularity can be recoverd. However, a simple change of the effective charge in the plasma cannot explain the observed levels of confinement improvement in the present models

Recent results on electron cyclotron current drive and MHD activity in RTP

The RTP tokamak (R = 0.72 m, a = 0.164 m, B-phi < 2 5.T, I-p = < 150 kA) is equipped with three gyrotrons (2 x 60 GHz, 180 kW, 100 ms each; 1 x 110 GHz, 500 kW, 200 ms) for electron cyclotron heating (ECH) and current drive (ECCD). The power from one of the 60 GHz gyrotrons is launched via an adjustable mirror from the high field side (HFS) in the 1X-mode. The power of both other gyrotrons is sent in perpendicularly to the toroidal magnetic field from the low field side (LFS). A comprehensive set of high-resolution multichannel plasma diagnostics is available to study the detailed behaviour of various plasma phenomena. First, recent diagnostic innovations are briefly discussed. Then, new physics results are presented for ohmic and EC heated plasmas. ECCD, slide-away discharges, discharges with a hollow temperature profile and MHD phenomena, including sawteeth and disruptions. are treated

Author Correction: Magnetic configuration effects on the Wendelstein 7-X stellarator

\u3cp\u3eIn the version of this Article originally published, and in the associated Publisher Correction, the members of the W7-X Team were not included. All versions of the Article, and the Publisher Correction, have now been amended to include these team members.\u3c/p\u3

Erratum to:magnetic configuration effects on the Wendelstein 7-X stellarator (Nature Physics, (2018), 14, 8, (855-860), 10.1038/s41567-018-0141-9)

\u3cp\u3eIn the version of this Article originally published, A. Mollén’s affiliation was incorrectly denoted as number 10; it should have been 1. Throughout the Article, some technical problems in typesetting meant that the tilde symbol above b and one instance of a superscript 2 were too high to be visible; see the correction notice for details. Finally, the citation to ref. \u3csup\u3e35\u3c/sup\u3e on page one of the Supplementary Information was incorrect; it should have been to ref. 36. These issues have now been corrected.\u3c/p\u3

Magnetic configuration effects on the Wendelstein 7-X stellarator

\u3cp\u3e The two leading concepts for confining high-temperature fusion plasmas are the tokamak and the stellarator. Tokamaks are rotationally symmetric and use a large plasma current to achieve confinement, whereas stellarators are non-axisymmetric and employ three-dimensionally shaped magnetic field coils to twist the field and confine the plasma. As a result, the magnetic field of a stellarator needs to be carefully designed to minimize the collisional transport arising from poorly confined particle orbits, which would otherwise cause excessive power losses at high plasma temperatures. In addition, this type of transport leads to the appearance of a net toroidal plasma current, the so-called bootstrap current. Here, we analyse results from the first experimental campaign of the Wendelstein 7-X stellarator, showing that its magnetic-field design allows good control of bootstrap currents and collisional transport. The energy confinement time is among the best ever achieved in stellarators, both in absolute figures (τ \u3csub\u3eE\u3c/sub\u3e > 100 ms) and relative to the stellarator confinement scaling. The bootstrap current responds as predicted to changes in the magnetic mirror ratio. These initial experiments confirm several theoretically predicted properties of Wendelstein 7-X plasmas, and already indicate consistency with optimization measures. \u3c/p\u3

Overview of first Wendelstein 7-X high-performance operation

\u3cp\u3eThe optimized superconducting stellarator device Wendelstein 7-X (with major radius R = 5.5 m, minor radius a = 0.5 m, and 30 m3 plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of 1-4.5 × 10\u3csup\u3e19\u3c/sup\u3e m\u3csup\u3e-3\u3c/sup\u3e with central electron temperatures 5-10 keV were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to 1.4 × 10\u3csup\u3e20\u3c/sup\u3e m\u3csup\u3e-3\u3c/sup\u3e were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of 8 · 10\u3csup\u3e19\u3c/sup\u3e m\u3csup\u3e-3\u3c/sup\u3e a temperature of 3.4 keV with Te/Ti = 1 was transiently accomplished, which corresponds to nTi(0)TE = 6.4 × 10\u3csup\u3e19\u3c/sup\u3e keV s m\u3csup\u3e-3\u3c/sup\u3e with a peak diamagnetic energy of 1.1 MJ and volume-averaged normalized plasma pressure {B}= 1.2%. The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above 1 × 1020 m\u3csup\u3e-2\u3c/sup\u3e line integrated density and Te = Ti = 2 keV central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.\u3c/p\u3