In the quest for new energy sources, the research on controlled thermonuclear fusion 1 has been boosted by the start of the construction phase of the International Thermonuclear Experimental Reactor (ITER). ITER is based on the tokamak magnetic configuration 3, which is the best performing one in terms of energy confinement. Alternative concepts are however actively researched, which in the long term could be considered for a second generation of reactors. Here, we show results concerning one of these configurations, the reversed-field pinch 4,5 (RFP). By increasing the plasma current, a spontaneous transition to a helical equilibrium occurs, with a change of magnetic topology. Partially conserved magnetic flux surfaces emerge within residual magnetic chaos, resulting in the onset of a transport barrier. This is a structural change and sheds new light on the potential of the RFP as the basis for a low-magnetic-field ohmic fusion reactor.The main magnetic field configurations studied for the confinement of toroidal fusion-relevant plasmas are the tokamak 3 , the stellarator 6 and the reversed-field pinch 4,5 (RFP). In the tokamak, a strong magnetic field is produced in the toroidal direction by a set of coils approximating a toroidal solenoid, and the poloidal field generated by a toroidal current flowing into the plasma gives the field lines a weak helical twist. This is the configuration that has been most studied and has achieved the best levels of energy confinement time. Thus, it is the natural choice for the International Thermonuclear Experimental Reactor, which has the mission of demonstrating the scientific and technical feasibility of controlled fusion with magnetic confinement.The RFP, like the tokamak, is axisymmetric and exploits the pinch effect due to a current flowing in a plasma embedded in a toroidal magnetic field. The main difference is that, for a given plasma current, the toroidal magnetic field in a RFP is one order of magnitude smaller than in a tokamak, and is mainly generated by currents flowing in the plasma itself. This feature is underlying the main potential advantage of the RFP as a reactor concept, namely the capability of achieving fusion conditions with ohmic heating only in a much simpler and compact device. In the past, this positive feature was overcome by the poorer stability properties, which led to the growth and saturation of several magnetohydrodynamic (MHD) instabilities, eventually downgrading the confinement performance. These instabilities, represented by Fourier modes in the poloidal and toroidal angles θ and φ as exp [i(mθ − nφ) were considered as an unavoidable ingredient of the dynamo self-organization process 4,8,9 , necessary for the sustainment of the configuration in time. The occurrence of several MHD modes resonating on different plasma layers gives rise to overlapping magnetic islands, which result in a chaotic region, extending over most of the plasma volume 10 , where the magnetic surfaces are destroyed and the confinement level is modest. This conditi...

With the exploration of the MA plasma current regime in up to 0.5 s long discharges, RFX-mod has opened new and very promising perspectives for the Reversed Field Pinch (RFP) magnetic configuration, and has made a significant progress in understanding and improving confinement and in controlling plasma stability. A big leap with respect to previous knowledge and expectations on RFP physics and performance has been made by RFX-mod since the last 2006 IAEA Fusion Energy Conference. A new self-organised helical equilibrium has been found (the Single Helical Axis-SHAx-state), which is the preferred state at high current. This regime is characterized by strong core electron transport barriers, with electron temperature gradients comparable to those achieved in tokamaks, and by a factor four improvement in confinement time with respect to the standard RFP. RFX-mod is also providing leading edge results on real-time feedback control of MHD instabilities, of general interest for the fusion community.

We describe experiments with a biased electrode inserted into the scrape-off layer (SOL) of the CASTOR tokamak. The resulting radial and poloidal electric field and plasma density modification are measured by means of Langmuir probe arrays with high temporal and spatial resolutions. Poloidally and radially localized stationary structures of the electric field (convective cells) are identified and a related significant modification of the particle transport in the SOL is observed.

This paper summarizes the main achievements of the RFX fusion science program in the period between the 2008 and 2010 IAEA Fusion Energy Conferences. RFX-mod is the largest reversed field pinch in the world, equipped with a system of 192 coils for active control of MHD stability. The discovery and understanding of helical states with electron internal transport barriers and core electron temperature >1.5 keV significantly advances the perspectives of the configuration. Optimized experiments with plasma current up to 1.8 MA have been realized, confirming positive scaling. The first evidence of edge transport barriers is presented. Progress has been made also in the control of firstwall properties and of density profiles, with initial first-wall lithization experiments. Micro-turbulence mechanisms such as ion temperature gradient and micro-tearing are discussed in the framework of understanding gradient-driven transport in low magnetic chaos helical regimes. Both tearing mode and resistive wall mode active control have been optimized and experimental data have been used to benchmark numerical codes. The RFX programme also provides important results for the fusion community and in particular for tokamaks and stellarators on feedback control of MHD stability and on three-dimensional physics. On the latter topic, the result of the application of stellarator codes to describe three-dimensional reversed field pinch physics will be presented.

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A novel diagnostic, the tunnel probe, is used to investigate the edge plasma of the CASTOR tokamak. Comparison with conventional small, cylindrical Langmuir probes (typical of those usually employed for turbulence measurements in magnetized plasmas) demonstrates the superiority of the tunnel probe. The collectors of the tunnel probe are concave, eliminating in theory all uncertainty of the effective collecting area, and thereby rendering the measurement of parallel ion current density, electron temperature, and floating potential more reliable. Two tunnel probes, mounted back-to-back in a Mach probe arrangement, are used to investigate directional asymmetries of the plasma parameters. The tunnel probes are used as standard Langmuir probes by applying voltage sweeps simultaneously to all their internal conductors. The measured electron temperature is higher on the electron side than on the ion side, and the floating potential is lower. The observed asymmetries, measured at low density and collisionality in CASTOR, could be consistent with a hot tail of non-thermal electrons flowing in the counter-current direction. The ratio of ion saturation currents to the internal conductors of the tunnel probe provides a second independent measurement of electron temperature whose directional asymmetry is less pronounced, in agreement with recent theoretical predictions that tunnel probes should be less sensitive to non-thermal electrons than Langmuir probes in certain conditions.

Experimental investigations of the plasma potential and electric field were performed for

The magnetized Mach probe is used to make measurement of plasma flows in the scrape-off layer of the Aditya tokamak [R. Jha et al., Plasma Phys. Controlled Fusion 51, 095010 (2009)]. This probe is further used to measure dependencies of Mach number on local plasma densities and radial distances of the probe in the scrape-off layer. The measured Mach number has contributions from E Â B drift, Pfrisch-Schl€ uter, and transport driven flows. We have determined that the toroidal flow is towards the ion side of the limiter and the poloidal flow direction is towards the contact of the last closed flux surface with the limiter. V C 2012 American Institute of Physics.