We discuss the processes underlying the excitation of fishbone-like internal kink instabilities driven by supra-thermal electrons generated experimentally by different means: Electron Cyclotron Resonance Heating (ECRH) and by Lower Hybrid (LH) power injection. The peculiarity and interest of exciting these electron fishbones by ECRH only or by LH only is also analyzed. Not only the mode stability is explained, but also the transition between steady state nonlinear oscillations to bursting (almost regular) pulsations, as observed in FTU, is interpreted in terms of the LH power input. These results are directly relevant to the investigation of trapped alpha particle interactions with low-frequency MHD modes in burning plasmas: in fact, alpha particles in reactor relevant conditions are characterized by small dimensionless orbits, similarly to electrons; the trapped particle bounce averaged dynamics, meanwhile, depends on energy and not mass.

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The existence of fishbone fluctuations at frequencies comparable to those of geodesic acoustic modes (GAM) and beta induced Alfvén eigenmodes (BAE) has been demonstrated theoretically in a recent work (Zonca et al 2007 Nucl. Fusion 47 1588). Here, we show that observation of fishbones at unexpectedly high frequencies in JET (Nabais et al 2005 Phys. Plasmas 12 102509) is well interpreted as experimental evidence of high (GAM/BAE range) frequency fishbones and discuss the insights concerning both supra-thermal particles as well as thermal plasma properties that can be obtained from experimental observations.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (B T up to 8.5 T) and plasma current (I p up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5-1 MeV. The total power input will be in the 30-40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P /R ∼ 22 MW m −1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ≈1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant

In this paper we present the fusion advanced studies torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of plasma-wall interaction, burning plasma physics, ITER relevant operation problems and steady state scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen in a wide range of parameters from high performance H-mode (toroidal field, B T , up to 8.5 T; plasma current, I P , up to 8 MA) to advanced tokamak (AT) operation (I P = 3 MA) as well as full non-inductive current scenario (I P = 2 MA). The coupled heating power is provided with 30 MW delivered by an ion cyclotron resonance heating system (30-90 MHz), 6 MW by a lower hybrid system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an electron cyclotron resonant heating system (170 GHz − B T = 6 T) for MHD and localized electron heating control and, eventually, with 10 MW by a negative neutral ion beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ion induced as well as turbulence fluctuation spectra, thus addressing the cross-scale couplings issue of micro-to meso-scale physics. The non-inductive scenario at I P = 2 MA is obtained with 60-70% of bootstrap current and the remaining by LHCD. Predictive simulations of the H-mode scenarios have been performed by means of the JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and shape control studies are also presented for the reference scenario.

Since the end of 2005 most of the plasma-wall interaction experiments on FTU have been focused on the possible use of liquid lithium as the plasma facing material. Liquid lithium limiter is an active method to deposit, during the plasma discharge, a lithium film on the walls with prolonged beneficial effects. Reliable operation with very clean plasmas, very low wall particle recycling, spontaneous peaking of the density profile for line-averaged density values n e > 1.0 × 10 20 m −3 have been obtained. These results have allowed us to extend the density limit to the highest value so far obtained (n e = 4.0 × 10 20 m −3 at I p = 0.7 MA and B T = 7.1 T, q a = 5.0, by gas puffing only) and to increase the energy confinement time by almost 50% with respect to the average value of 50 ms of the old ohmic FTU database. An accurate analysis of these plasmas has been carried out by means of a gyrokinetic code to establish the role of collisionality and density gradients on the observed phenomenology.

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 turbulence in the scrape-off layer (SOL) plasma of FTU is characterized in order to assess its effect on the current drive efficiency of the lower hybrid (LH) waves. Amplitude, frequency and perpendicular wave vector of the fluctuations are measured for a variety of the main plasma conditions in front of the LH antenna together with the temperature and density in the SOL and used as inputs for the linear scattering theory of the LH waves developed many years ago. This theoretical model can account for both the frequency spectral broadening of the LH pump and the variations of the driven current, inferred by the perpendicular fast electron bremsstrahlung signals. The fraction of the LH power that is then deduced to be effective for current drive appears to be well related to the calculated optical thickness τ of the SOL. It drops as low as 40% as τ increases, consistent with the model prediction. Possible ways to control the SOL optical depth are investigated and a clear relation of the fluctuation level with the collisionality is found.