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.

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

Joint European Torus (JET) plasma initiations that form a significant quantity of runaway electrons have been studied. It is shown that there is no direct relationship between the prefill pressure and breakdown electric field and signs of runaway electrons during the plasma initiation. Runaway electron generation is determined by the electric field and density development at and after burn-through. A clear criterion of density and electric field at one given point in time, which would ensure the avoidance of runaway electron generation, cannot be determined, because the timescales for the formation of runaway electrons and for the dynamics of the density differ significantly. Moreover, the formation process can be reversed, reducing the influence of runaway electrons on the discharge. Ensuring a high enough density will reduce the likelihood that runaway electron discharges are formed. It is also found that at JET the electric field often exceeds the critical electric field during the early stages of the current ramp-up phase, even when no signs of runaway electrons are present. Expected runaway current dynamics have been analysed using the discharge circuit equation. The comparison of the expected runaway electron current dynamics shows it to be significantly slower compared to theoretical expectations in the presence of a hot and dense thermal background plasma. This could be explained by an enhanced critical electric field and/or a reduced confinement of runaway electrons. The latter is shown to be affected by bursts of magnetohydrodynamic activity that are characteristic during the current ramp-up. The development of discharges in which the current is fully carried by runaway electrons happens on a slow timescale of several seconds, limited by the available flux. Such timescales are sufficient for improved active control of these events, avoiding runaway currents at plasma initiation exceeding values at which they could damage in-vessel components. The results provide insight into the improvement and interpretation of self-consistent modelling of runaway electron generation during the start-up of International Thermonuclear Experimental Reactor discharges.

New experiments have been conducted at DIII-D to improve the physics understanding of plasma initiation assisted by Electron Cyclotron (EC) wave injection, allowing better extrapolation to ITER. This has been achieved by applying an EC pulse prior to start of the inductive plasma initiation (i.e. the generation of a loop voltage). A pre-plasma was formed during the EC pulse that was characterized in terms of the maximum density and temperature. Parametric scans were performed to study the influence of the EC injected power, EC injection angle, and pre-fill gas pressure on the pre-plasma creation process. These experiments showed that pre-ionized plasma of good quality can have a significant effect on the subsequent Vloop induced plasma initiation process, i.e. a high density pre-plasma, increases the plasma current rise and speed at which ionization is achieved when the Vloop is applied. A good quality pre-plasma is one that achieved a significant degree of ionization, mainly obtained by providing sufficient ECH power in DIII-D of the order of 1 MW. It was found that a minimum EC power of 0.5 MW was required in DIII-D to create ionization, and this would scale to a minimum power of roughly 6.5 MW for ITER.

Ion temperatures of over 100 million degrees Kelvin (8.6keV) have been produced in the ST40 compact high-field spherical tokamak (ST). Ion temperatures in excess of 5keV have not previously been reached in any ST and have only been obtained in much larger devices with substantially more plasma heating power. The corresponding fusion triple product is calculated to be ni0Ti0τE≈6±2×1018m-3keVs. These results demonstrate for the first time that ion temperatures relevant for commercial magnetic confinement fusion can be obtained in a compact high-field spherical tokamak and bode well for fusion power plants based on the high-field ST.