[en] Stationary operations have been achieved at JET in ITBs scenarios, with the discharge time limited only by plant constraints. Full current drive was obtained, all over the high performance phase, with the current density profile frozen by using Lower Hybrid current drive. For the first time a feed-back control on the total pressure and on the electron temperature profile was implemented by using respectively the Neutral Beams and the Ion Cyclotron waves. Although impurity accumulation could be a problem in steady state ITBs, these experiments bring some elements to answer to it. Tokamak operation in enhanced confinement regimes, characterized by edge and/or Internal Transport Barriers (respectively known as H-mode and ITB), is attractive as it represents an important step towards the approach of ignition conditions. Moreover, the necessity of steady state operation in a Tokamak reactor, has led to the concept of the Advanced Tokamak, in which the current density profile is no longer tied to the plasma conductivity and is non inductively driven. Since the bootstrap current is a consequence of the pressure gradient, one of the primary goal of the Advanced Tokamak studies is to maximize the bootstrap fraction, with a proper alignment, both in H mode and in ITB regimes. However, for several reasons, it is difficult to envisage an operational situation in which the bootstrap fraction is close to 100%: for instance, there are few chances of pressure or/and current profile control to optimize the MHD stability. So far, various experiments have been performed with improved confinement regimes lasting up to tens of the confinement time and up to some current relaxation times. In some experiments a large non inductive plasma current (< 75%) was obtained with about 50% from bootstrap and 25% from Neutral Beam Injection (NBI); however, no full current drive operation was achieved and, moreover, with the available heating systems, no active feedback control of the current density and of the plasma pressure profiles was performed. In other experiments quasi-steady state regimes were achieved by operation at high poloidal beta, with a large fraction of bootstrap current and by using the current drive capability of the negative NBI. Finally stationary discharges (∼70 s) in full current drive were achieved in Tore Supra by using LHCD (Lower Hybrid Current Drive). In principle, the simultaneous use of several auxiliary power systems makes possible to sustain high performance regimes using an active control of all the different plasma profiles. The availability of three heating systems (NBI, ICRH (Ion Cyclotron Heating) and LHCD) gives JbT some advantage with respect to other large experiments, allowing, for the first time, two important targets of advanced scenario to be achieved simultaneously: a) plasma configurations with both electron and ion ITBs have been obtained in full current dnve (no inductively driven current) by optimizing the coupling of the LHCD system, the duration of these discharges being constrained only by plant limitations (the toroidal magnetic field flat top) (author)

[en] Full text: Advanced Tokamak scenarios include two different regimes: 'steady state' (characterized by the presence Internal Transport Barrier: ITB) and the 'hybrid' scenario. So far both the regimes have always been obtained in presence of strong injection of external momentum by Neutrals Beam Injection (NBI) heating. By using Lower Hybrid Current Drive (LHCD) to sustain the central q slightly above one and with a large plasma region having the magnetic shear close to zero, an 'hybrid scenario' has been established, for the first time, in discharges with dominant Ion Cyclotron Resonance Heating (ICRH) and with a normalized beta close to two. By starting from a configuration with reversed magnetic shear (sustained only by LHCD) and with a well established ITB on the electron specie, an ITB also on the ions specie has been obtained by using ICRH in an ion heating scheme, (3He )D. No external momentum input was provided by the NBI, except for the charge-exchange and the MSE beams. In these discharges the evaluated ExB shearing rate was always very small and lower than analytical evaluations of the turbulence growth rate. (author)

[en] 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, in burning plasma relevant conditions, fast particle physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than present day experiments by 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 (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study fast particle behaviours in 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, 4MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative beams (NNBI) in the energy range of 0.5-1 MeV. The total power input will be in the 30-40 MW range in the different plasma scenarios with a wall power load comparable with that of ITER (P/R∼22 MW/m). 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 ≅ 2.5), to a full current drive scenario, lasting around 170 s. The first wall as well as the divertor plates will be of Tungsten in order to ensure reactor relevant operation regimes. The divertor itself is designed to be completely removable by remote handling. This will allow studying (in view of DEMO) the behaviour of innovative divertor concepts, such as those based on liquid Lithium. FAST is capable of operations with very long pulses, up to 170 s, despite that it is a copper machine. The magnets initial operation temperature is 30 K, with cooling realised by helium gas. The in vessel components, namely first wall and divertor, are actively cooled by pressurised water at 80 0^C. The same water is also used to back up the vacuum vessel. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple down to 0.3%

[en] In the FTU machine there are the following coils: main transformer coil (T), preprogrammed vertical field coil (V) and feedback coil (F). In order to maximize the number of the ports for diagnostics apparatus and additional RF heating, FTU has been designed without a massive copper shell. Plasma displacements due to preprogramming failures or disruptions, will be recovered by the vertical field generated by the coil. In addition, also the plasma current is feedback controlled by regulating the transformer current. To study the control laws and to optimize the regulation gains, a computer code based on the CSMP III (continuous system modeling program) has been provided. 4 refs., 6 figs

[en] Full text: The data from the Hybrid scenario experiments has been incorporated into a database that comprises ∼100 discharges and ∼280 time points, and contains more than 30 plasma and scenario parameters. Analysis of the database has revealed that there is a spectrum of temperature gradient scale lengths that exceed the critical level considered at JET to indicate the presence of an ITB. In the best example (B_T = 2.5 T, I_p = 2.1 MA, P_NBI 13.4 MW, P_ICRH = 2.5 MW, 1 < q(0) < 1.5) it was R/L_Ti ∼ 14, R/L_Te ∼ 11) on a wide region (r ∼ 0.2 m) of the plasma core (r/a ∼ 0.4). These steep gradients were steady for almost 5 seconds. The central ion and electron temperature were T_i(0) ∼ 17 keV and T_e(0) ∼ 8 keV. In the initial phase of the heating pulse the edge temperatures had been quite low (T_i ∼ T_e ∼ 2 keV), but suddenly they increased to ∼4 keV while, at the same time, the edge density decreased providing a more or less constant edge pressure. The discharge was essentially without ELMs, although the additional heating power was well above both the usual H-mode power threshold. The global performance was comparable to, or slightly better than, that of an equivalent standard Hybrid discharge with type I ELMs (β_N ∼ 2, H₈₉ ∼ 2.2). The plasma energy associated with the pedestal was W_ped ∼ 37% in this discharge, whilst the part within the steep gradient region was W_core ∼ 23%. Despite the relatively low density the toroidal rotation was only about half that of a typical JET ITB plasma. Nevertheless, the E x B shearing rate is calculated to be very large. Analysis using the gyrokinetic code, KINEZERO shows that, in the improved confinement region, ITG-TEM wavelength instabilities should be stable. An important factor in the plasmas with improved core confinement appears to be the density, which was higher in the case of similar Hybrid plasmas without core confinement improvement. Transport and stability analysis of discharges with and without core confinement improvement will be presented and compared. So far plasmas with improved core confinement in the Hybrid domain of JET represent a small minority of the database and occur only at low density and plasma collisionality. However, the possibility to obtain good global confinement in plasmas without significant ELM activity would be a useful extension of the Hybrid scenario operational domain and very attractive for ITER operations. (author)

[en] A computer program for analysing the absolute and relative stabilities of any complex system by the root-locus method was developed. It is used to reanalyse the present horizontal position feed-back control in the ASDEX tokamak and to select the optimum parameters for this system with RCL filters for reducing thyristor noise. (orig.)

[en] The Volt-second consumption in FTU has been evaluated both during the plasma start-up and during the current ramp-up from a flat top level to a higher flat top level. The Poynting definition was used to work out the resistive volt-second consumption; the equilibrium reconstruction ODIN code was used to get the physical quantities necessary (β_p+l_i/2, β_p, l_i, V_p, L_e). The Ejima parameter E=ΔΦ_surface/μ₀RI_p, where ΔΦ_surface is the flux consumption inside the last magnetic surface, R is the major radius, I_p is the plasma current, has been evaluated for a large set of shots in H₂, D₂ and He. The database covers a large variation of plasma parameters: I_p=.35:1.1 MA; dI_p/dt∼2:9MA/s; n_e∼1.5:10x10^-19 m^-3; for the whole database it was B_T=6T,R=.93m; a=.28m, b/a∼1.04. The flux necessary to get the plasma break-down can be estimated in 0.1 volt-second for the first 50 kA of the plasma current and it is included in all the computations. The total flux consumption (internal plus external) obtained from the equilibrium code outputs was cross-checked with the total flux measured by the poloidal circuit currents variation: the two results were quite well in agreement (about 5%). The mean Ejima parameter obtained for the examined shots was E∼1.15. (author) 6 refs., 3 figs

[en] We have investigated the magnetic signals of the FT tokamak (R=83 cm, a_L=20 cm) at the plasma edge using two coils to detect B_θ placed 3.5 cm apart from each other in the shadow of the poloidal limiter at the external upper side of the torus. The measured frequency response of the two 4-winding coils, 1 cm long and radius 0.45 cm, shows a roll-off of -3 dB at about 300 kHz, well above the range of frequency considered (<200 kHz). The electronic chain is composed of two amplifiers with an optical decoupler between them to isolate the coils, short-circuited with the liner, from the control room devices. The signals were sampled typically at a frequency of 2 MHz for a time window of 8 ms and were then integrated numerically to obtain B_θ. (author) 5 refs., 8 figs

[en] The possibility of designing stellarator or torsatron configuration at low aspect ratio remains an ambitious goal. The advantages of such an approach are the high use of the magnetic field volume, the reduced space taken by the windings and the computed equilibrium and ideal stability high beta limit, whereas the disadvantages are the increased fast particle losses. The choice of the winding law for the helicoidal coils is certainly the most critical tool for maximizing the advantages and minimizing the disadvantages. The choice of winding the helicoidal currents on circular cross section tori with constant pitch in toroidal coordinates (θ,ω,φ) (θ=const; mω+nφ=const., that means a higher pitch on the outside of the torus and a lower one on the inside) has turned out to be a rewarding one; it allows a fully analytical treatment of the 3-D vacuum fields and permits an easy access to low aspect ratio configurations. In this paper we attempt a systematic scan in the aspect ratio of idealized stellarators, with fixed poloidal m=2 number and toroidal field B^T₀=5T at R=R_o=2m (concentration point of the toroidal coordinates). (author) 7 refs., 7 figs., 1 tab

[en] The use of fully toroidal co-ordinates permits the two-dimensional problem of the axisymmetric plasma toroidal equilibrium to be reduced to the one-dimensional problem of determining a limited number of its toroidal multipolar moments. This has allowed the creation of a fast semi-analytic predictive equilibrium code that can be used in both free and fixed boundary conditions for plasmas with circular or mildly non-circular cross-section. The concept of toroidal multipoles is also particularly suitable for the analysis of experimental data from magnetic probe measurements and clarifies the conditions under which the plasma thermal and electrical self-inductances βsub(p) and lsub(i) can be estimated separately. Finally, the interpretation of the magnetic equilibrium measurements in terms of toroidal multipoles can directly provide the boundary conditions for a fast equilibrium reconstruction code. Examples of the application of such a code to the JET magnetic measurements are reported. (author)