The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device’s unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power ‘starvation’ reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in–out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added.

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Pellet injection has been used as a primary fueling scheme in Large Helical Device (LHD). Pellet injection has extended an operational region of NBI plasmas to higher densities with maintaining preferable dependence of energy confinement on density, and achieved several important data, such as plasma stored energy (0.88 MJ), energy confinement time (0.3 s), β (2.4 % at 1.3 T) and density (1.1×10 20 m-3). These parameters cannot be attained by gas puffing. Ablation and subsequent behavior of plasma has been investigated. Measured pellet penetration depth that is estimated by duration of the Hα emission is shallower than predicted penetration depth from the simple neutral gas shielding (NGS) model. The penetration depth can be explained by NGS model with fast ion effect on the ablation. Just after ablation, redistribution of ablated pellet mass was observed in short time (~ 400 µs). The redistribution causes shallow deposition and low fueling efficiency.

Recent large helical device experiments revealed that the transition from ion root to electron root occurred for the first time in neutral-beam-heated discharges, where no nonthermal electrons exist. The measured values of the radial electric field were found to be in qualitative agreement with those estimated by neoclassical theory. A clear reduction of ion thermal diffusivity was observed after the mode transition from ion root to electron root as predicted by neoclassical theory when the neoclassical ion loss is more dominant than the anomalous ion loss. Neoclassical ion transport is important in stellarator plasmas, because the helical ripple losses are comparable to or sometimes even higher than the anomalous losses in contrast to those in tokamaks. The crucial issues of neoclassical ion transport are (1) the reduction of ion thermal diffusivity due to a large positive radial electric field in the electron root [1][2][3], and (2) the reduction of ion thermal diffusivity due to the optimization of the magnetic field structure (s optimization) [3][4][5][6]. However, there has been no experimental study to test these issues on the neoclassical ion transport in stellarator plasmas. This is because the transition of the radial electric field from small negative (ion root) to large positive (electron root) was observed only in plasmas with the assistance of electron cyclotron heating (ECH), where electron heating is dominant [7][8][9][10][11]. The ion temperature is much lower than the electron temperature because ions are heated only by the energy exchange between ions and electrons. In these experiments, the significant increase of electron temperature and a clear reduction of electron thermal diffusivity were observed in the plasma core in the electron root. However, no reduction of ion thermal diffusivity was observed because of the lack of direct ion heating. There have been no experimental results to show the improvement of ion transport in the electron root, although a significant improvement of ion transport (rather than the electron transport) is predicted by the neoclassical theory [6]. This paper describes the experimental results of the neoclassical feature of ion transport for the first time, the reduction of ion thermal diffusivity due to the transition to the large positive electric field (electron root), and/or the optimization of the magnetic field structure (s optimization).The large helical device (LHD) [12] is a Heliotron device (poloidal period number L 2, and toroidal period number M 10) with a major radius of R ax 3.5 4.1 m, an average minor radius of 0.6 m, and magnetic field up to 3 T. The radial electric field ͑E r ͒ is derived from the poloidal and toroidal rotation velocity and pressure gradient of neon impurity measured with charge exchange spectroscopy [13] at the midplane in LHD (vertically elongated cross section) using a radial force balance. The radial force balance equation can be expressed as E r ͑en I Z I ͒ 21 ͑≠p I ͞≠r͒ 2 ͑y u B f 2 y f B u ͒, where B f and B u are toroidal a...