A sharp transport barrier, accompanied by a bifurcated poloidal rotation and a radial electric field, is formed at the plasma edge by driving a radial current across the outer magnetic surfaces of a tokamak. A decrease in particle transport is observed for negative radial E fields. When the radial current is turned off, the E field and the rotation damp on a time scale comparable with the ion-ion collision time.PACS numbers: 52.25.Fi, 52.55.Pi, 52.70.Ds An intensive effort is in progress, worldwide, to correlate the large and growing data base of experimental observations on plasma transport in magnetic fusion containment devices with theoretical models. In particular, the so-called "//^-mode" regimes in tokamaks have attracted strong interest due to their enhanced particle and energy confinement properties. ^""^ The transition of a plasma into the H mode is marked by a sudden decrease in the hydrogenic light emission from the plasma edge, followed by a prolonged increase in the plasma density. The reduction of hydrogen light (H^ or H^) indicates that the incoming neutral particle flux is reduced, presumably because of a decrease of the outgoing plasma flux, leading to a reduction in "recycling." The improvement in the energy confinement is generally less than the increase in particle confinement, ^-mode measurements also reveal the formation of sharp density and temperature gradients inside the last closed magnetic surfaces, which represents a transport barrier. Despite the magnitude of the effort aimed at modeling the H mode, no clear mechanism has been identified, although radial electric fields are thought to play a role."*'^ In this Letter, experimental observations confirming the importance of the radial E field and the associated plasma rotation for 7/-mode confinement are presented.In 1979, electron injection was used to modify the edge potentials in order to reduce ion sputtering in the Macrotor tokamak.^ Subsequently, improved particle confinement and a concomitant impurity accumulation were observed,^ apparently giving rise to an H mode. These effects were attributed to the creation of edge radial electric fields and associated negative plasma potentials much larger in magnitude than Tgia), where a is the plasma radius. Recently we have extended this earlier work using the new, titanium-gettered. Continuous Current Tokamak (CCT) at the University of California, Los Angeles. The recent experiments clearly show the //-mode signatures found in other tokamaks in various limiter, divertor, and auxiliary-heating configurations. The previously seen impurity limitations^ have also been eased by new electrode designs. ^ For the //-mode-regime studies, CCT was operated in the pulsed neo-Alcator regime, with central parameters R^l.5 m, a =0,4 m, Bt=3 kG, /p=50 kA, ne=5 xlO*Vcm^ Kioop<1.5 V, Te>l50 eV, and T/> 100 e = 180 INSULATOR ELECTRODE = 0" C/2 Q 1.5 1.8 MAJOR RADIUS (M) FIG. 1. (a) Cross section of tokamak, a ^40 cm, showing the location of the exciting electrode, re ^25 cm, and the "rake" probe arrays used...

Demonstrating improved confinement of energetic ions is one of the key goals of the Wendelstein 7-X (W7-X) stellarator. In the past campaigns, measuring confined fast ions has proven to be challenging. Future deuterium campaigns would open up the option of using fusion-produced neutrons to indirectly observe confined fast ions. There are two neutron populations: 2.45 MeV neutrons from thermonuclear and beam-target fusion, and 14.1 MeV neutrons from DT reactions between tritium fusion products and bulk deuterium. The 14.1 MeV neutron signal can be measured using a scintillating fiber neutron detector, whereas the overall neutron rate is monitored by common radiation safety detectors, for instance fission chambers. The fusion rates are dependent on the slowing-down distribution of the deuterium and tritium ions, which in turn depend on the magnetic configuration via fast ion orbits. In this work, we investigate the effect of magnetic configuration on neutron production rates in W7-X. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from DD fusion and, particularly, on the 14.1 MeV neutron production rates. Despite triton losses of up to 50 %, the amount of 14.1 MeV neutrons produced might be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3  ×  1019 m−3, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.