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A tokamak plasma with internal transport barriers (ITBs) is the best candidate for a steady ITER operation, since the high energy confinement allows working at plasma currents (Ip) lower than the reference scenario. To build and sustain an ITB at the ITER high density (1020 m−3) and largely dominant electron (e−) heating is not trivial in most existing tokamaks. FTU can instead meet both requests, thanks to its radiofrequency heating systems, lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC up to 1.2 MW). By the combined use of them, ITBs are obtained up to peak densities ne0 > 1.3 × 1020 m−3, with central e− temperatures Te0 ≈ 5.5 keV, and are sustained for as long as the heating pulse is applied (>35 confinement times, τE). At ne0 ≈ 0.8 × 1020 m−3Te0 can be larger than 11 keV. Almost full current drive (CD) and an overall good steadiness is attained within about one τE, 20 times faster than the ohmic current relaxation time. The ITB extends over a central region with an almost flat or slightly reversed q profile and qmin ≈ 1.3 that is fully sustained by off-axis lower hybrid current drive. Consequent to this is the beneficial good alignment of the bootstrap current, generated by the ITB large pressure gradients, with the LH driven current. Reflectometry shows a clear change in the turbulence close to the ITB radius, consistent with the reduced e− transport. Ions (i+) are significantly heated via collisions, but thermal equilibrium with electrons cannot be attained since the e−–i+ equipartition time is always 4–5 times longer than τE. No degradation of the overall ion transport, rather a reduction of the i+ heat diffusivity, is observed inside the ITB. The global confinement has been improved up to 1.6 times over the scaling predictions. The ITB radius can be controlled by adjusting the LH power deposition profile that is affected mostly by the q value of the discharge, while the ITB strength can be varied through central EC heating. FTU experiments have shown that ITER-like e-ITBs are achievable.

MHD spectroscopy in FTU has revealed high frequency (HF) oscillations between 30 and 70 kHz that accompany the development of large m0 = −2, n0 = −1 islands in ohmic plasmas. This frequency range is one order of magnitude above the island rotation frequency, one order of magnitude below the first toroidal gap in the Alfvén continuum and of the same order of the low frequency gap introduced by finite beta effects. The HF spectrum is organized in pairs of modes with opposite toroidal numbers, n = 1 and n = −1 for the stronger ones, which form standing waves in the island rest frame. The poloidal structure of HF modes is |m| = 2 for the main lines. HF lines appear when the level of poloidal field perturbation produced by the island at the plasma edge exceeds 0.2%. The absence of energetic ions in FTU ohmic plasmas, the strict correlation between HF and tearing modes and the existence of a threshold indicate that HF modes tap their energy from the tearing mode by a non-linear coupling mechanism.

An overview of the FTU results during the period 2003–4 is presented. A prototype ITER-relevant lower hybrid current drive (LHCD) launcher, the passive active multijunction, has been successfully tested (f = 8 GHz) showing high power handling and good coupling properties and current drive comparable to those of a conventional launcher. Effective electron and ion heating (via collisions) have been achieved with the 140 GHz ECRH systems up to 1.5 MW, as well as current drive (25 kA at ne0 = 8 × 1019 m−3). The mitigation of disruptions has been studied with on-axis ECRH. Ion Bernstein wave studies have shown the importance of recycling in achieving improved confinement plasmas. Advanced tokamak scenarios are presented including repetitive pellet enhanced plasmas and electron internal transport barriers (e-ITBs). Very peaked density profiles have been achieved with a low speed vertical pellet injector located at about mid-radius on the high field side. The performance is comparable to that achieved with a high-speed horizontal pellet injector. Possible reasons for this behaviour are discussed, among them the presence of an ‘MHD’ drift once particles reach the q = 1 surface. The effect of peaked density profiles on confinement is discussed. Electron ITBs can be produced at high density in FTU with LHCD only and with the combined use of LHCD and ECRH: Te0 = 6 keV with ne0 = 1.4 × 1020 m−3 and H97 = 1.6. Turbulence is strongly reduced. Ions are heated by collisions with ΔTi/Ti up to 35% showing that e-ITBs are not degraded by the electron–ion collisions. Particle pinch studies have been made at high densities in full current drive conditions where the Ware pinch plays no role. An anomalous inward pinch exists even at these high densities (ne0 = 1.5 × 1020 m−3). Despite the absence of energetic particles in FTU, MHD spectroscopy has revealed high frequency modes (30–80 kHz) that might have consequences for burning plasmas.

A ‘hybrid’ scenario for ITER is defined through its objectives: a large fusion yield for a long time duration. In many tokamaks, discharges characterized by a stationary current density profile, enclosing a large volume of low magnetic shear with q0 near 1, have achieved improved confinement and higher beta limits. Their extrapolation to ITER from existing data corresponds to the ITER hybrid scenario. These discharges are characterized by soft MHD events. Physics issues relevant to the existence and extrapolation of this scenario will be addressed. New JET experiments with a large component of RF heating have answered some of these issues: injected momentum is not essential, hybrid scenarios are achievable at low ρ*, hybrid regimes have been achieved with ITER-relevant Te/Ti and they are compatible with a very low edge activity/low pressure pedestal. Data from pure RF discharges in other tokamaks (FTU, TS, TCV) seem to confirm that a large volume of low magnetic shear with q0 close to 1 is the key to achieving hybrid scenarios. Issues needing resolution in the extrapolation to ITER are discussed. The present understanding provides encouraging prospects for the use of this scenario in ITER.

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E×B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E×B shear rate is comparable to the linear growth rates at the location of the ITB.