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Si tratta della prima parte di una ampia antologia prampoliniana che si comporrà di tre volumi (anche del secondo è ormai prossima l’uscita). I tre curatori ripartono da una analoga iniziativa editoriale che venne intrapresa circa trent’anni fa da Renzo Barazzoni e Nelson Ruini, arricchendola però in modo consistente, tanto da delineare i contorni di una nuova opera. Di quel lavoro del 1981 viene riproposta, tra le altre cose, la breve presentazione firmata da Gaetano Arfè.

The first tokamak discharges with deuterium-tritium mixtures have been carried out in the Joint European Torus (JET). The main objectives were to produce more than IMW of fusion power in a controlled way, to determine tritium retention in torus systems and to establish effective means of tritium removal. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. Deuterium plasmas were heated by high power deuterium neutral beams from fourteen sources and fuelled by two neural beam sources injecting tritium. In the best (D-T discharge), the tritium concentration was about 11% at peak performance, when total neutron emission rate was 6.0*1017 s-1, with 1.7 MW of fusion power. The fusion amplification factor QDT was 0.15. With optimum tritium concentration, this pulse would have produced a fusion power approximately=5 MW and nominal QDT=0.46. The same extrapolation for the best pure deuterium discharge gives about 11 MW and a nominal QDT=1.14. Techniques for introducing, tracking, monitoring and recovering tritium were highly effective.

Transport studies are presented in this chapter. Global scaling studies have been performed using several transport codes. Ohmic plasmas are found to follow the ITER97 L-mode scaling. Transport coefficients are discussed for improved confinement scenarios achieved in the Frascati Tokamak Upgrade (FTU): the repetitive pellet enhanced plasma mode, showing neoclassical confinement with H-factors up to 1.6, and the electron internal transport barriers (ITBs) with large transport barriers and H-factors up to 1.3. Heat transport models have been tested using electron cyclotron resonance heating (ECRH), steady or modulated, as a probe. The electron temperature stiffness observed in the main bulk of steady FTU plasmas can be interpreted both with a critical gradient transport model and with a model based on the existence of canonical profiles. ECRH has also been used to benefit from the improved confinement generally associated with low or negative magnetic shear, and large electron temperatures have been achieved in these conditions. Profile resiliency is observed so that heat transport is not consistent with a constant thermal diffusivity. Experimental optimization is discussed together with the analysis of transport coefficients. Thorough discussions of impurity transport are given, both for intrinsic and injected (from laser blow-off) impurities. Code simulation and experimental data are compared for a series of FTU experiments focusing on the improved confinement modes (pellets and ITBs). A moderate inward pinch velocity is generally required to reproduce the data.

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Experiments performed on FTU tokamak, aiming at validation of physics-based transport models of the electron temperature profile resilience, are presented. ECRH is used to probe transport features, both in steady-state and in response to perturbations, while ECCD and LHCD are used for current density profile shaping. Observed confinement behaviour shows agreement with a critical temperature gradient length modelling. Central, low gradient plasma is characterized by low stiffness and low electron thermal diffusivity. Strong stiffness and high conduction are found in the confinement region. Resilience is experimentally characterized by an index of the resistance of the profile to adapt its shape to localized ECRH, while the diffusivity and its low-high transition are measured both by power balance and heat pulse propagation analysis. A particular attention is given to the investigation of the transition layer between low-high diffusivity and low-high stiffness regions. A dependence of LTc on magnetic shear, similar to what found in Tore Supra, and consistent with ETG based anomalous transport, is found.

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Gaussian beam steering (2 beams, *400 kW each) and toroidal field tuning are used to control ECRH localization for MHD stabilization and energy transport studies on FTU tokamak. It is shown that by locating the absorption of one beam inside the m=2 island, and of a second one close to the s.t. inversion radius, both instabilities can be suppressed, with a significant impact on energy confinement. MHD suppression allows the analysis of the underlying energy transport. It is shown that the electron temperature and pressure profiles are stiff when ECRH is localized off-axis, and that a finite amount of heat pumping against the temperature gradient must be accounted for to explain steady-state profiles at high density (^//>JC,«1020m"3). The resistance of the thermal response to change with the heating source term can be analyzed in terms of two models implying profile stiffness: the effect of a critical temperature gradient, possibly related to ETG modes, and the tendency of a tokamak plasma to stick to preferred profiles. In particular, it is shown that the assumption of constant magnetic entropy, implying relaxation of the toroidal current density, allows good reproduction of the experimental profiles and of the main observation of a step-wise radial profile of the effective electron thermal diffusivity. Profile stiffness is observed both at high density, with significant e-i coupling, and at low density when much less power is conducted through ions.

An important focus of present research on thermonuclear fusion based on magnetic confinement is the control of rotating magnetic islands of low helical mode numbers m,n that are the nonlinear stage of resistive instabilities developing at magnetic surfaces where q=m/n. These perturbations are often driven by external resonant "error fields" or mode coupling effects and maintained by local distortions of the current profile. The islands can be reduced in width or completely suppressed by a current driven by electron cyclotron waves (ECCD) accurately located within the island. A requisite for an effective control action is the ability of identifying the relevant state variables (radial location, island amplitude, frequency and phase) in "real time" and vary accordingly the control variables (wave beam power modulation and direction). Traditionally Magnetohtydrodynamic (MHD) instabilities are monitored by external magnetic signals, that however give no information about the radial location of the perturbation within the plasma. Here we describe the design of a real time diagnostics/control device using an original algorithm for processing Electron Cyclotron Emission signals to monitor temperature fluctuations and EC wave power deposition.

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Experiments performed on the FTU tokamak, aimed at validation of physics-based transport models, which yield the electron temperature profile stiffness by using electron cyclotron resonance heating (ECRH), and at the implementation of ECRH-based techniques for plasma real-time control, are presented. ECRH is used to probe transport features, both in the steady state and in response to time-varied heating. The experiments clearly show stiffness in the electron temperature profile response to localized ECRH. The low gradient plasma region near the axis is characterized by low stiffness, in the sense that the temperature gradient can change with increasing heat flux, and low electron thermal diffusivity. Strong stiffness (in the sense that the temperature gradient length 1/LT =− ∇ T/ T does not change significantly even with significantly different heating profiles and intensity) and high diffusivity are found in the confinement region (0.15 < r/a < 0.5). Particular attention is given to the experimental investigation of the transition layer between low and high diffusivity (and low and high stiffness) regions, which is located at the EC deposition radius rdep when powerful ECRH is applied. A transition layer, identified by heat waves launched by modulated ECH, is found also in plasmas with dominant Ohmic heating, showing that it is a local plasma feature, and not merely a consequence of a step-wise increase in the conducted heat flux at rdep. All observations fit well with a critical temperature gradient length modelling of local electron heat transport. The measurement of the rate of change of stored electron energy when ECRH is switched on (or off) is the basis of a technique for real-time and automatic detection of the absorption layer. Repetitive short pulses are used to increase the signal to noise ratio, with a very low duty cycle, so as not to waste ECRH power during the detection procedure.

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The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to produce more than one megawatt of fusion power in a controlled way; (ii) to validate transport codes and provide a basis for accurately predicting the performance of deuterium-tritium plasma from measurements made in deuterium plasmas; (iii) to determine tritium retention in the torus systems and to establish the effectiveness of discharge cleaning techniques for tritium removal; (iv) to demonstrate the technology related to tritium usage; and (v) to establish safe procedures for handling tritium in compliance with the regulatory requirements. A single-null X-point magnetic configuration, diverted onto the upper carbon target, with reversed toroidal magnetic field was chosen. Deuterium plasmas were heated by high power, long duration deuterium neutral beams from fourteen sources and fuelled also by up to two neutral beam sources injecting tritium. The results from three of these high performance hot ion H-mode discharges are described: a high performance pure deuterium discharge; a deuterium-tritium discharge with a 1% mixture of tritium fed to one neutral beam source; and a deuterium-tritium discharge with 100% tritium fed to two neutral beam sources. The TRANSP code was used to check the internal consistency of the measured data and to determine the origin of the measured neutron fluxes. In the best deuterium-tritium discharge, the tritium concentration was about 11% at the time of peak performance, when the total neutron emission rate was 6.0 x 10(17) neutrons/s. The integrated total neutron yield over the high power phase, which lasted about 2 s, was 7.2 x 10(17) neutrons, with an accuracy of +/- 7%. The actual fusion amplification factor, Q(DT), was about 0.15. With an optimum tritium concentration, this pulse would have produced a fusion power of almost-equal-to 5 MW and a nominal Q(DT) almost-equal-to 0.46. The same extrapolation for the pure deuterium discharge would have given almost-equal-to 11 MW and a nominal Q(DT) = 1.14, so that the total fusion power (neutrons and alpha-particles) would have exceeded the total losses in the equivalent deuterium-tritium discharge in these transient conditions. Techniques for introducing, tracking, monitoring and recovering tritium were demonstrated to be highly effective: essentially all of the tritium introduced into the neutral beam system and, so far, about two thirds of that introduced into the torus have been recovered.