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This article presents a new method for estimating the electron temperature of the Protosphera's screw pinch. The temperature radial profile is obtained by a self-consistent modeling of a 1D MHD equilibrium along with a 0D power balance of the plasma column, given measurements and estimates of the axial pinch plasma current, of the plasma rotational frequency and, at the equatorial plane, of the electron density radial profile, of the edge poloidal magnetic field, of the edge electron temperature and of the neutrals pressure in the vacuum vessel. The plasma is considered in equilibrium with its neutral phase and in constant rotation. A MATLAB code has been developed with the aim of estimating the MHD radial equilibrium profiles, the thermodynamic plasma state and the neutrals profile. The numerical estimates are compared with available experimental data showing a good agreement., Comment: 4 pages, 6 figures, 1 table, research presented to the "6th ICFDT"

The onset of magnetic reconnection in space, astrophysical and laboratory plasmas is reviewed discussing results from theory, numerical simulations and observations. After a brief introduction on magnetic reconnection and approach to the question of onset, we first discuss recent theoretical models and numerical simulations, followed by observations of reconnection and its effects in space and astrophysical plasmas from satellites and ground-based detectors, as well as measurements of reconnection in laboratory plasma experiments. Mechanisms allowing reconnection spanning from collisional resistivity to kinetic effects as well as partial ionization are described, providing a description valid over a wide range of plasma parameters, and therefore applicable in principle to many different astrophysical and laboratory environments. Finally, we summarize the implications of reconnection onset physics for plasma dynamics throughout the Universe and illustrate how capturing the dynamics correctly is important to understanding particle acceleration. The goal of this review is to give a view on the present status of this topic and future interesting investigations, offering a unified approach.

Analytical solutions of the Grad–Shafranov equilibrium equation in simply connected plasma configurations, comprised of toroidal magnetic surfaces and open surfaces connected to electrodes, are reviewed and generalised. The Grad–Shafranov equation is linearised introducing assumptions on plasma current and pressure, which preserve regularity of solutions on the symmetry axis, as required for a simply connected geometry. Particular solutions are found by separation of variables both in cylindrical coordinates and in spherical ones. Equilibria that model local or global features of PROTO-SPHERA plasmas are constructed by combining a few particular solutions.

PROTO-SPHERA is a magnetic confinement device for controlled fusion experiments which forms a plasma spherical torus not around a metal centerpost (as in spherical tokamaks), but around a plasma centerpost. Thanks to a vacuum vessel in polymethylmethacrylate (PMMA) transparent to visible light, a system of six fast visible light cameras has been implemented around the PROTO-SPHERA experiment so as to perform an optical tomography. To do so, synchronous triggered frames acquisition and exposure time control have been performed using the MARTe2 software framework. Also, a proper calibration/reconstruction of each line of sight for all the cameras has been implemented, so as to compute a 3D representation of the vessel. A 2D tomography inversion of the plasma column equatorial plane has been performed under the consideration that the value of each pixel represents the integrated brightness along its line of sight. The camera acquisition system, the experimental setup and the obtained results will be presented.

We analyze the temperature distribution on the vertical axis of an axisymmetrical plasma, the radial temperature being almost constant. We solve the equation for the temperature analytically. Further we investigate the bidimensional flow inside the plasma, showing the existence of a convective cell. The considered plasma is the central column of the PROTO-SPHERA configuration, which aims at forming a confined plasma torus around the central plasma discharge.

A simple steady-state model for a 3-species mixture (ions, electrons, and neutrals) in a screw-pinch plasma configuration is developed. The model is applied to the central plasma column of the PROTO-SPHERA experiment. Degree of ionization, azimuthal current density, and azimuthal ion velocity are calculated. Full ionization is found at plasma temperatures above 1.5 eV, with neutrals confined in an outer shell where radial plasma flow develops and drives both azimuthal current and azimuthal flow.

PROTO-SPHERA (Spherical Plasma for Helicity Relaxation Assessment) is a new concept of torus that aims to produce a spherical torus with closed flux surfaces and a force-free Screw Pinch with open flux surfaces, fed by electrodes. In order to verify the stability of the configuration and the technical components of the PROTO-SPHERA initial arc, the initial Phase-1 experiment was built, and its plasma, in the year 2108, has reached both the total current (10 kA) and total duration (1 s) targets. A two colour interferometer is one of the diagnostic system proposed to provide PROTO-SPHERA with the measurement of its electron density, which is currently of the order of ∼1020 m−3. It will allow to obtain line density, necessary to characterize this uncommon plasma, as integrate measurement along the equatorial plane. The design and the interferometer characteristics are described in this paper.

Plasma properties can be useful in a wide spectrum of applications. Experimental projects on controlled nuclear fusion are the most challenging of these applications and, at the same time, the best way to approach plasma science. Since nuclear fusion reactors can ensure a large-scale, safe, environmentally-friendly and virtually inexhaustible source of energy, several fusion-oriented megaprojects and innovative companies are appearing all over the world. PROTO-SPHERA (Spherical Plasma for HElicity Relaxation Assessment) is the first plasma project with a simply connected configuration, namely not requiring additional objects inside the plasma volume. This is obtained by a plasma arc, shaped as a screw pinch, acting as the centerpost of a spherical torus with minimal aspect ratio. Due to its intrinsic physical, engineering and economic advantages, this new approach is attractive also on an industrial scale and with several developments that still needs to be explored. This paper presents the PROTO-SPHERA basic principles, its first encouraging results and its expected and potential evolutions.

PROTO-SPHERA is the first plasma project with a simply connected configuration, namely not requiring other objects inside the plasma. This is obtained by a hydrogen plasma arc, shaped as a screw pinch, acting as the central column of a spherical torus with minimal aspect ratio. The reduced size and complexity of the configuration can lead to several physical and engineering advantages, providing relevant contributions to the magnetic plasma confinement methods. Moreover, due to the limited costs and requirements, the set-up is attractive for replication and developments also on an industrial scale with many potential applications. This paper describes the initial activities of the PROTO-SPHERA project, with emphasis on the electrical tests and commissioning performed to implement the plasma central column. In particular, an intermediated experiment classified as MULTI-PINCH was designed to investigate the breakdown conditions and the pinch stability. © 2015 IEEE.

The design study of PROTO-SPHERA, a novel compact torus configuration, has been completed. It is composed of a spherical torus (ST) (with closed flux surfaces) and a force-free screw pinch (SP) (with open flux surfaces and fed by electrodes). PROTO-SPHERA is formed at spherical-tokamak-like densities (∼10 19 m −3 ) with low voltage (∼200 V) between the electrodes. The idea of replacing the metal centrepost current (Itf ) of the spherical tokamaks with the SP plasma electrode current (Ie) is aimed mainly at getting rid of the rod at the centre of the plasma configuration, which is the most critical component of spherical tokamak design. As a consequence it should be possible to decrease the aspect ratio A = R/a (R = ST major radius, a = ST minor radius) in the course of experiment and to increase the ratio between the toroidal plasma current (IST) and the plasma electrode current, IST/Ie � 1. Matching two plasma configurations, i.e. an open flux-surface SP and a closed flux-surface ST, brings to life several radically new issues. The purpose of this paper is to analyse the equilibrium, the ideal MHD stability and the formations and modelling issues of such a combined magnetic confinement system. The MULTI-PINCH experimental setup, which is being assembled inside the START vacuum vessel (now in Frascati), will represent the first phase of PROTO-SPHERA: its goal is to prove the feasibility of a stable disc-shaped SP around the electrodes.

PROTO-SPHERA is a proposed spherical torus where a hydrogen plasma arc, in a form of a screw pinch field fed by electrodes, replaces the central conductor. This simply connected magnetic configuration, if fusion relevant, might strongly simplify the design of a fusion reactor. The machine design philosophy, basic geometry and operating conditions together with the major components like the vacuum vessel, coils, electrodes, protection components, divertor, etc. are analyzed. The thermal and electromagnetic behavior as well as the predicted and permitted key stresses will be discussed in order to demonstrate that the design, construction and reliable operation of the machine are feasible. Reference is also made to the proposed Multi-Pinch experiment using the START vacuum vessel to demonstrate the feasibility and stability of the PROTO-SPHERA configuration.

When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper.European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference.

High density plasmas (n0 ≈ 8 × 1020m-3) featuring steady improved core confinement have been obtained in FTU up to the maximum nominal toroidal field (8 T) by deep multiple pellet injection. These plasmas also feature high purity efficient electron-ion coupling and peaked density profiles sustained for several confinement times. Neutron yields in excess of 1 × 1013 n/s are measured, consistent with the reduction of the ion transport to neoclassical levels.

Cryogenic pellet injection has been used in START and it is under commissioning for MAST for fuelling and diagnostics. In START NBI, heated plasmas fuelled with deuterium pellets, high normalized density (Greenwald number NG≤1.1) with good energy confinement time (τE/τEelmfITER97≤1.3) both at high beta (βT≤25%, βN≤4.9) have been obtained for time scales close to the energy confinement time (∼τE). In addition, high fuelling efficiency (∼100%) have also been simultaneously observed. A complete pellet injection programme is currently under development in MAST, including the commissioning of a eight pellet injector, technical aspects of launching pellets through guide tubes, dedicated diagnostics, and modelling.

The ignition experiment Ignitor will produce high performance plasma regimes with neutron fluxes at the first wall comparable to those expected in power generating reactors (1015n/cm2/s at the first wall). An appreciable degradation of the inorganic insulators surrounding the conductors is expected and the measurement of basic plasma parameters, such as plasma current and position, by electromagnetic diagnostics can become problematic. Thus an R&D program, aimed at the development of effective and affordable devices for electromagnetic diagnostics with higher damage threshold, is being carried out. Prototype diagnostic coils of different shapes have been manufactured, in collaboration with SALENTEC and Università di Lecce. At the same time, an alternative plasma position control method is being explored, based on the diffraction and detection of the soft X-ray radiation emitted, mostly by Molybdenum impurities, near the top or bottom of the plasma column, where the distance of the plasma last closed magnetic surface from the wall is only few millimeters. The underlying principles and general layout of the new system is essentially an adaptation of a 2-dimensional curved crystal spectrometer. © 2008 American Institute of Physics.

In toroidal magnetic confinement configurations all the charged particles align their magnetic drifts with the flux surfaces if, and only if, the lines of force have zero geodesic curvature (omnigeneous configuration). In this condition, the neoclassical transport is zero; unfortunately, standard tokamak devices are far from meeting such a condition. Nevertheless, a minimization of the geodesic curvature of the lines of force in tokamaks minimizes the neoclassical transport, and could as well reduce any other transport terms depending on the noncoincidence between magnetic and drift surfaces. In particular, the geodesic curvature of the lines of force generates both the neoclassical Pfirsch–Schluter current density j∥PS, as well as Hirshman factor q [Nucl. Fusion 18, 917 (1978)], which increases the moment of inertia of the magnetic configuration. The improved confinement observed in the VH-mode (very high confinement) and in reversed magnetic shear discharges of the DIII-D tokamak [Plasma Physics Contro...

Electron temperatures in excess of 8 keV have been obtained by electron-cyclotron-resonance heating on FTU plasmas at peak densities up to $8\ifmmode\times\else\texttimes\fi{}{10}^{19}\mathrm{m}{}^{\ensuremath{-}3}$. The magnetic shear in the plasma core is low or negative, and the electron heat diffusivity remains at, or below, the Ohmic level $(0.2\mathrm{m}{}^{2}/\mathrm{s})$, in spite of the very large heating power density $(10--20\mathrm{MW}/\mathrm{m}{}^{3})$ which produces extremely high temperature gradients (up to 120 keV/m). The ion heat transport remains at the neoclassical level.

Magnetic measurements alone on spherical tori should allow a very good separation of the poloidal beta beta p, from the internal self-inductance li/2 and should even permit an accurate estimate of the current density jphi profile. However, the reduced space allowed for magnetic sensors near the central conductor in a spherical tokamak, and the possibility of producing flux core spheromak configurations without a central conductor, could imply that magnetic probes are not present in the cavity of the spherical torus. The fluxes and fields of a variety of calculated spherical torus configurations, all endowed with a single or double null separatrix, are analysed in terms of spherical multipolar moments obtained from simulated magnetic measurements located only upon a sphere surrounding the spherical plasma. The solution to the problem of the absence of magnetic measurements in the cavity of the spherical torus is to fix from non-magnetic measurements (e.g., spectroscopy) the plasma inboard boundary rin on the equatorial plane. This constraint is added to the constraints of matching the spherical multipolar expansion in an iterative solution of the Grad-Shafranov equation, on the basis of a spherical geometry. The convergence of the spherical reconstructive equilibrium code is extremely fast and gives an error on the total plasma current Ip of less than 1% at an aspect ratio A=1.2, an error on the position of the plasma boundary of less than 2% of the radius of the plasma sphere, an error on beta p of at most 15% and, finally, the jphi profile is extremely well reconstructed in peaked, flat and even hollow cases. The effect of an uncertainty +or- delta rin upon the spectroscopic identification of the plasma inboard boundary on the equatorial plane rin is assessed

Over the last few years, a great deal of effort has been devoted to solving the problem of power and particle handling in divertors, which has been recognized as a critical issue for the operation of a magnetic fusion reactor. In particular, the choice of materials for plasma facing components has been examined with a view to developing heat and erosion resistant materials for divertor target plates. A large database on the behaviour of low-Z (carbon or beryllium) materials in tokamaks is available, while for high-Z materials there is little experience in the present generation of magnetic fusion devices. Frascati Tokamak Upgrade (FTU), a high field compact tokamak, has devoted part of its experimental campaign to studying the plasma characteristics when its limiter material is changed from the usual Inconel (nickel) to molybdenum and tungsten. Siliconization of the machine has also allowed the comparison of plasma performance when a relatively low-Z (silicon) ion is the dominant impurity. In this article, results are reported concerning the plasma operation, the differences in plasma characteristics and radiation losses, the impurity generation mechanisms and the relative impurity concentrations in the core plasma. A simulation of the experimental results, made with a self-consistent edge-core coupled model is presented, in order to provide evidence of the main physics mechanisms responsible for the observed behaviour

The authors report on their experience of treating ureteral stones in 1183 patients with ballistic and electrohydraulic endoscopic lithotripsy. Since the outcome was successful in 95% of cases with few complications (2%), this method is considered highly effective for treating stones in any part of the ureter, with the possibility of resolving renal colic in real-time. In the push-up cases (25% for sub-junctional ureteral lithiasis, 10% other sites) a second treatment was performed with ESWL after positioning a double J catheter.

A new analytical technique for extracting the Boozer magnetic coordinates in axisymmetric magnetohydrodynamic equilibria is described. The method is based upon the correspondence between the expansion of the flux function in toroidal multipolar moments and the expansion in toroidal axisymmetric harmonics of the magnetic scalar potential χ0, which appears in the covariant representation B=∇χ0+β ∇ψT of the magnetic field. An example of calculation of Boozer magnetic coordinates is given for an experimental highly shaped high β equilibrium of DIIID [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159].

The behaviour of the FTU tokamak plasma has been analysed by using two reconstructive MHD equilibrium codes: the first code works by using the magnetic data alone and the second one by including as well the shape of the kinetic pressure profile, as obtained from the measured profiles of electron temperature Te and density ne. The code that analyses the magnetic data alone provides a good evaluation of the macroscopic quantities such as the poloidal beta βp and the internal inductance li, if the plasma elongation is greater than 1.04. No detailed information about the toroidal current density profile J and the safety factor profile q can be obtained from the magnetic data alone. On the other hand, the coupling of magnetic and kinetic data is able to provide a reasonable estimate of the toroidal current density profile and of its behaviour during the plasma discharge. The reliability of the J and q profile reconstruction has been explored and validated by a detailed comparison with the observed MHD behaviour of the FTU plasma discharges. A good agreement between the appearance of the sawtooth activity and the drop of the safety factor on the magnetic axis go to unity is observed. Also, at least for edge safety factors qψ less than 4, the sawtooth inversion radius is found to be very close to the q=1 surface. A remarkable correspondence between J and Te3/2 is found in sawtoothing discharges. The structure of the snake oscillation in pellet injected discharges is found to be strictly correlated to the position of the q=1 surface. A cylindrical linear tearing mode stability calculation applied to the reconstructed J profile has shown qualitative agreement with the appearance of the Mirnov oscillations. Finally, the magnetic reconnections between double resonant surfaces during the rise of the plasma current or after the pellet injection have provided an interesting validation of the J profile reconstruction

In this paper, the properties of alpha-particle confinement of l = 2 stellarators are studied as a function of the aspect ratio (3 [lt] R/a [lt] 12). A collisionless orbit calculation has been performed numerically for stellarator configurations obtained by winding the helical currents on circular-cross-section tori with a constant pitch in toroidal coordinates. All the configurations studied exibit pronounced separatrixlike features that also rotate with constant pitch in toroidal coordinates. The fraction of alpha particles contained within the last closed magnetic surface rapidly increases with the aspect ratio and is [gt]80% for R/a [gt] 4. The escaping alpha particles cross the separatrix within narrow helical strips around the X-point path when it is between the inboard and the top of the torus. The particle motion is also followed outside the last closed magnetic surface. When the boundary (i.e., the surface where the helical currents flow) is sufficiently far from the plasma, the escaping alpha particles are found to remain trapped in the X-point region.

An exact prediction of the plasma contour is important for the design of new tokamaks and for the definition of the operational scenarios of existing machines. Certain features of the plasma boundary, such as elongation and triangularity, and a detailed description of the X-point region are relevant for the plasma-limiter interaction. The toroidal multipolar moment method has been used, with fixed boundary conditions, to produce the required plasma shape. The resulting vacuum flux function has been reproduced, using the poloidal coil currents that minimize the magnetic energy as best parameters for the fitting. Examples are given for the single and double X-point configurations in JET and ITER.

High Z materials, as possible plasma facing components for a fusion reactor, are being re-evaluated due to their advantages connected to the low sputtering yield and small erosion. A data base of experimental results related to their use on present Tokamak is missing. FTU has carried out a series of experiments with different medium and high Z limiter materials, Ni, Mo, and W, to improve the knowledge of the plasma behavior in these conditions. The results on plasma operation, core parameters, impurity content, and corresponding radiation losses are presented, and compared with those obtained with a low Z dominant impurity (siliconization). A simple model coupling the SOL and the plasma core is able to reproduce the experimental results, indicating that sputtering and self-sputtering are the main physical mechanisms of impurity production. © 1995 American Institute of Physics.

(1) The main impurities in FT are oxygen and iron. Iron impurities are negligible at high density (nFe < 10−4 ne). Radiation losses account for 20-40% of the total input power in all conditions [1, 2].(2) FT is provided with a complete poloidal limiter made of stainless steel. The same material is used for the vacuum chamber. To a good approximation, the volume inside the outermost closed flux surface equals 2π2Rgeoab.

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The capability of the poloidal field coil system, as presently designed, to provide an effective vertical stabilization of the plasma in the IGNITOR machine has been investigated using the CREATE-L response model. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The growth rate of the vertical instability and the power required by the active stabilization system has been estimated with this model. The possible failure of the relevant electromagnetic diagnostics has been taken into account, evaluating the robustness of the plasma position reconstruction strategy. A realistic description of the power supply system has permitted to carry out the optimization of the proportional-integrative-derivative (PID) controller, both with a voltage and a current loop control scheme. An assessment of the requirements for the plasma cross section shape control has been carried out considering perturbations of the plasma global parameters independent of each other and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been rated relative to their capability to restore shape modifications due to different plasma disturbances. The most effective coil combination, that minimizes recovery time and voltage required, has been identified. (C) 2007 Elsevier B.V. All rights reserved.

PROTO-SPHERA is a proposed spherical torus (ST) where a hydrogen plasma arc, in a form of a Screw Pinch (SP), fed by electrodes, replaces the central rod of the standard ST experiments. This machine, with a longitudinal pinch current Ie=60 kA, will produce an ST of diameter 2Rsph=75 cm, aspect ratio A=1.2-1.3, carrying a toroidal current Ip=120-240 kA. The plasma will be magnetically shaped as a disk near each modular annular electrode. Theoretical calculations (using a 3D ideal MHD stability code) show that such a configuration should be ideally stable up to a total beta ranging between 15-25%, depending upon the ratio Ip/Ie. The ST toroidal current should be sustained by helicity injection from the Screw Pinch. In order to demonstrate the feasibility of the SP discharge and the basic electrode characteristics, a preliminary linear electrode test bed, Proto-Pinch, has been built and operated successfully. The detail mechanical engineering design of PROTO-SPHERA has now been completed. The machine design philosophy, basic geometry and operating conditions with the major components like the vacuum vessel (VV), coils, electrodes, protection components and divertor will be analysed. The thermal and electromagnetic behaviour as well as the predicted and permitted heat stresses will be discussed in order to demonstrate that the design, construction and reliable operation of the machine are feasible. Reference is also made to the proposed Multi-Pinch experiment using the START VV to demonstrate the feasibility and stability of the PROTO-SPHERA configuration. The START VV has now been transported and disassembled in ENEA-Frascati. The detailed design of Multi-Pinch, initial phase of PROTO-SPHERA, has also been completed and the first contract related to the PF coils has been placed

— The authors report on the treatment of kidney staghorn stones by a transureteral approach. Endoscopic debulking can be followed by ESWL. The methods, results and advantages versus other techniques (PUL) are evaluated.

— The Authors report on two cases of inadvertently transplanted renal carcinomas and on a five-year follow-up after conservative surgery. They discuss the ethical-legal questions connected with this treatment option.

Simply connected Chandrasekhar‐Kendall‐Furth plasma configurations (CKF) contain a magnetic separatrix, dividing a spherical torus (ST), two secondary tori and a surrounding discharge. Axisymmetric CKF equilibria are calculated imposing a constant relaxation parameter μ=μ0j⋅B/B2 at the edge of the plasma and a surface averaged 〈μ〉 decreasing toward the axis of the ST. A poloidal current flows in the surrounding discharge and injects magnetic helicity into the ST. ∇p is supposed to be concentrated in the same region where ∇〈μ〉 has the largest variation. CKF equilibria are stable to all ideal MHD perturbations with low toroidal mode numbers (n=1,2,3), up to β=2μ0〈p〉Vol/〈B2〉Vol≈1, even in absence of a nearby conducting shell. The poloidal current driven in the surrounding discharge by different orbits of co/counter‐circulating promptly lost fusion products could help in sustaining a CKF reactor.

The EC system at 140 Ghz, 2 MW, being implemented on FTU tokamak for performing electron heating, profile control and current drive, is composed by four gyrotrons with 0.5 s pulse length capability, and four hybrid mirror/waveguide transmission lines. The launching system is capable of poloidal/toroidal beam steering through a set of tiltable in-vacuum mirrors, coupling to the O-mode at the fundamental resonance, with full control of the e.m. field polarization. Experiments have been made with two gyrotrons, by launching 0.8 MW to the plasma. In the high density regime (local n{sub e} from {approx_equal}0.8 up to 2 10{sup 20} m{sup -3}) the e-i energy transfer is significant, and ion heating is observed through the enhancement of the neutron emission in Deuterium plasmas. The highest electron heating is observed when on-axis ECRH is performed in the current ramp-up, sawtooth-free phase. In steady-state conditions, saw-tooth period is increased up to stabilization by off-axis ECRH. Localized ECCD is also performed to assist in the shaping of the current density profile.

About 400 kW of rf power at the frequency of 433 MHz were coupled to the plasma of the FTU tokamak by a waveguide antenna. During the experiment, a peaking of the Te and ne profiles and a Hα drop have been observed during the IBW pulse, only when operating with toroidal magnetic field value which allows the wave penetration to in the plasma bulk. Moreover, the electron temperature increased typically from about 2.5 keV to about 4.5 keV with a coupled power of about 350 kW; the line-averaged central plasma density increased from 3×1019 m−3 to about 5×1019 m−3.

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Results are reported of the 8 GHz Lower Hybrid experiments on FTU after the installation of the new toroidal limiter. A figure of merit of the Current Drive efficiency eta(CD) approximate to 0.11.10(20) A/Wm(2) is estimated for plasma density = 10(20) m(-3) and no appreciable broadening of the launched frequency is detected. In low density experiments sawteeth are stabilised and m = 1 activity is present in the plasma. Shear reversed discharges with large reversal radius, r(S)/a approximate to 0.5, are obtained at higher density, lower temperature, B-T = 4 T, q(a) approximate to 5.5, by off-axis LH CD. The reversed configurations exhibit high central temperature coexisting with regular m = 2, n = 1 relaxations of large amplitude and are maintained up to LH switch off. At higher magnetic field, B = 5.2 T, q(a) approximate to 7, irregular DTM crashes are present during the whole LH pulse. Confinement time of radiofrequency heated discharges (P-LH = 0.5 divided by 2.P-OH) exhibits the same behaviour of FTU ohmic discharges following the ITER89-P scaling. Preliminary results of central 140 GHz Electron Cyclotron Resonant Heating (ECRH) during the plasma current ramp-up, aimed at obtaining shear reversed configurations are also reported.