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 nonaxisymmetric 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 minimise 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 (E > 100ms) 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 W7-X plasmas, and already indicate consistency with optimisation measures.
The Magnetized Plasma Interaction Experiment (MAGPIE) is a versatile helicon source plasma device operating in a magnetic hill configuration designed to support a broad range of research activity and is the first stage of the Materials Diagnostic Facility at the Australian National University. Various material targets can be introduced to study a range of plasma-material interaction phenomena.Initially, with up to 2.1 kW of RF at 13.56 MHz, argon (10 18 -10 19 m −3 ) and hydrogen (up to 10 19 m −3 at 20 kW) plasma with electron temperature ∼3-5 eV was produced in magnetic fields up to ∼0.19 T. For high mirror ratio we observe the formation of a bright blue core in argon above a threshold RF power of 0.8 kW. Magnetic probe measurements show a clear m = +1 wave field, with wavelength smaller than or comparable to the antenna length above and below this threshold, respectively. Spectroscopic studies indicate ion temperatures <1 eV, azimuthal flow speeds of ∼1 km s −1 and axial flow near the ion sound speed.
An extensive examination of the plasma response to dominantly n = 2 non-axisymmetric magnetic perturbations (MPs) on the DIII-D tokamak shows the potential to control 3D field interactions by varying the poloidal spectrum of the radial magnetic field. The plasma response is calculated as a function of the applied magnetic field structure and plasma parameters, using the linear magnetohydrodynamic code MARS-F (Liu et al 2000 Phys. Plasmas 7 3681). The ideal, single fluid plasma response is decomposed into two main components: a local pitch-resonant response occurring at rational magnetic flux surfaces, and a global kink response. The efficiency with which the field couples to the total plasma response is determined by the safety factor and the structure of the applied field. In many cases, control of the applied field has a more significant effect than control of plasma parameters, which is of particular interest since it can be modified at will throughout a shot to achieve a desired effect. The presence of toroidal harmonics, other than the dominant n = 2 component, is examined revealing a significant n = 4 component in the perturbations applied by the DIII-D MP coils; however, modeling shows the plasma responses to n = 4 perturbations are substantially smaller than the dominant n = 2 responses in most situations.
Classical particle drifts are known to have substantial impacts on fluxes of particles and heat through the edge plasmas in both tokamaks and stellarators. Here we present results from the first dedicated investigation of drift effects in the W7-X stellarator. By comparing similar plasma discharges conducted with a forward- and reverse-directed magnetic field, the impacts of drifts could be isolated through the observation of up-down asymmetries in flux profiles on the divertor targets. In low-density plasmas, the radial locations of the strike lines (i.e. peaks in the target heat flux profiles) exhibited discrepancies of up to 3 cm that reversed upon magnetic field reversal. In addition, asymmetric heat loads were observed in regions of the target that are shadowed by other targets from parallel flux from the core plasma. A comparison of these asymmetric features with the footprints of key topological regions of the edge magnetic field on the divertor suggests that the main driver of the asymmetries at low density is poloidal E × B drift due to radial electric fields in the scrape-off layer and private flux region. In higher-density plasmas, upper and lower targets collected non-ambipolar currents with opposite signs that also inverted upon field reversal. Overall, in these experiments, almost all up-down asymmetry could be attributed to the field reversal and, therefore, field-dependent drifts.
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