[en] The FT-Upgrade Tokamak in order to accomplish its aims has to operate its magnet under high stresses for a large number of shots. In order to assess the feasibility of a magnet capable of this performance, a three-dimensional finite element stress analysis has been carried out and a material cryogenic test program has been started. The main results of the stress analysis and the preliminary experimental results are illustrated. 5 refs

[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] The successful development of ITER and DEMO scenarios requires a preparatory activity on devices smaller than ITER, with sufficient flexibility and capable of investigating the peculiar physics of burning plasma conditions. The aim of the Fusion Advanced Studies Torus (FAST) proposal is showing that preparation of ITER scenarios and development of new expertise for DEMO design and R and D can be effectively implemented on a new facility that: a) will work with Deuterium plasmas, avoiding the problems associated with the use of Tritium, and will investigate non linear dynamics that are relevant for the understanding of alpha particle behaviors in burning plasmas by using fast ions accelerated by heating and current drive systems; b) will work in a dimensionless parameter range close to that of ITER; c) will test technical innovative solutions for the first wall/divertor directly relevant for ITER and DEMO, such as full-tungsten plasma facing components and advanced liquid metal divertor target; d) will exploit advanced regimes with long pulse duration with respect to the current diffusion time; e) will provide a test bed for ITER and DEMO diagnostics; all fast particles expected in FAST plasmas (p, D, ³He) are assumed to be diagnosed; as in ITER, diagnostics of fast particles require intensive R and D, that can be carried out on FAST with reduced costs and development time; f) will provide an ideal framework for model and numerical code benchmarks, verification and validation in ITER and DEMO relevant plasma conditions. The scientific rationale of the FAST conceptual design will be discussed. The choice of high equilibrium magnetic field B = 7.5 T, the consequent possibility of operating routinely at high plasma current I_p = 6.5 MA, high plasma density and moderate temperature, while maintaining a significant fusion performance at Q > 1 together with the possibility of extended pulse operations (up to 80 resistive times) will elucidate the capability for FAST of reaching its objectives in a flexible and cost-effective way. (author)

[en] FAST is the conceptual design for a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at integrated investigations of fast particle physics, plasma operations and plasma wall interaction in burning plasma relevant conditions. In Deuterium plasma operations, FAST has the capability to simultaneously approach relevant dimensionless physical parameters in all the ITER scenarios. The necessity of achieving ITER relevant power densities and performance with moderate cost has led to a compact Tokamak design (R=1.82 m , a= 0.64 m), with a high toroidal field (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study fast particle behaviours with dimensionless parameters similar to ITER, the project is based on a dominant Ion Cyclotron Resonance Heating system (ICRH; 30 MW coupled to the plasma). Moreover, the experiment foresees 6 MW of Lower Hybrid (LH), 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. Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI) in the energy range of 0.5-1 MeV. The total power input is 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 ITER scenarios can be studied: starting from the reference H-mode, with plasma edge and ELMs characteristics similar to those of ITER (Q up to ∼ 2.5), and arriving to full non-inductive current drive scenarios lasting ∼ 160 s, Under these conditions, first wall as well as divertor plates will be made of tungsten. 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 foreseeing the use of liquid lithium. FAST is capable to operate with very long pulses, up to 170 s, despite being a copper machine. The magnets initial operation temperature is 30 K and the cooling is ensured 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 for vacuum vessel baking. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple as low as 0.3%

[en] Separated First Wall Panels mechanically attached to a shield block is now the reference concept for the Primary Wall Modules of RTO/RC ITER. The objective of the present work is to demonstrate the practical feasibility of a First Wall Panel utilizing a duplex round (steel) in square (copper) heat sink wound around a steel core and covered by Beryllium armour tiles. These three different materials (Be, Cu, steel) are joined together by diffusion bonding. The Copper alloy/stainless steel and Copper alloy/Beryllium joints were studied and developed selecting the optimal parameters for the related diffusion process. Several specimens were manufactured to be mechanically and thermally tested. The joints were mechanically tested using dedicated press equipment and investigated by micro-structural analysis with optical and SEM microscopy. Some thermal tests were finally carried out using an Electron Beam Facility. A dedicated R and D programme has led to the development of a co-drawing process, suitable for manufacturing the duplex Copper alloy-stainless steel heat sink. Two mock-ups were manufactured, the first in reduced-scale to test the thermal performance of the system, the second of larger scale and geometry better to represent the First Wall Panel

[en] The TJ-2 HELIAC device is a stellarator proposed for the construction in Madrid by the Spanish Euratom-Ciemat Association. It is a medium size HELIAC axis stellarator with four periods, a major radius of 1.5 m, a toroidal magnetic field up to 1T, and a duty cycle up to 1 s every 5 min. The parameters of the TJ-2 machine are given. The vacuum vessel (VV) is an independent component fixed on the machine supports. It is formed to follow the shape of plasma. The VV is composed of 32 sectors and 32 rings, which are made of AISI 304 LN stainless steel to assure very low magnetic permeability. The construction of the VV is outlined. The method of suspending the VV with four suports is explained. Because of the symmetry, a three-dimensional FEM model was made for a quarter of the whole structure. The vacuum vessel element model, the results of the linear static analysis and the instability analysis are reported. Under the operating conditions, the mechanical behavior of the VV is fully satisfactory for both stress and displacement requirements. (K.I.)

[en] The basic idea in the design of the FTU (Frascati Tokamak Upgrade) machine is to try to reach a range of plasma parameters of great thermonuclear interest within a reasonable financial effort. In FTU, the good energy confinement properties are combined with the strong plasma heating obtained injecting up to 8 MW of radiofrequency power in the lower hybrid electron mode (8 GHz). The main engineering features of the toroidal machine, the remote handling tools for cutting and welding the vacuum vessel and the machine assembly and installation procedures are illustrated. (author)a. 3 refs.; 4 figs.; 1 tab

[en] This paper present the results of a stability analysis on a sector of the FTU (Frascati Tokamak Upgrade) toroidal vacuum vessel. FTU is an experimental machine, now under construction, mainly devoted to the study of the effects of lower hybrid radiofrequency heating on plasmas in reactor relevant conditions. Its vacuum chamber is a stainless steel structure completely welded with a major radius of 0.935 m and a minor radius of 0.335 m consisting of 12 toroidal thin sector joined together by thick ribs. This structure is loaded by comprensive electromagnetic forces both in toroidal and in radial direction that can create instability conditions. One of the major problems in its design is to determine the points where this phenomenon is likely to occur in order to avoid dangerous buckling situations. Theoretical analysis of the stability behaviour of one of these sector has been conducted by means of the ABAQUAS finite element code. The critical load has been determined by a classical algorithm and by the modified Riks methods. Both methods have given similar results in an elastic analysis. Furthermore the second one has been applied also using an elastoplastic model of the material to determine the critical load and the post buckling behaviour of the structure. Experimental tests have been conducted on a full scale model of the toroidal sector. The model has been placed in a large tank filled with water where the pressure has been gradually increased up to the collapse of the structure. The theoretical and experimental results have been compared and a good agreement has been found between them

[en] The industrialized countries most active in nuclear fusion (China, Korea, Japan, India, Russia, the United States and the European Union) are participating in the development of the ITER, an experimental reactor representing a fundamental step toward nuclear fusion energy. Europe, and Italy in particular, are deeply involved in this initiative with an important related project too.

[en] The shielding of the heat flux coming from plasma is one of the most limiting factors in plasma-facing component (PFC) design. In fact, the performance of a cooled divertor plate system is mainly limited by the heat transfer capability (maximum value of the critical heat flux CHF) and by the capability to sustain thermal stresses, even if the maximum allowable heat flux is determined by the thermal conductivity of the protective material (maximum temperature value on the plasma facing surface). A new concept for cooled thermal shield design was devised and tested. Analyses and tests demonstrate that the new concept introduces very high improvement in PFC design, in terms of both heat removal capability (very high CHF) and related stress performance. Up to 80 MW/m² under steady state were successfully applied