[en] FAST (Fusion Advanced Studies Torus), the Italian proposal of a satellite facility to ITER, is a compact tokamak (R₀ = 1.82 m, a = 0.64 m, triangularity δ = 0.4) able to investigate non linear dynamics effects of α-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO. Currently, ENEA is investigating the feasibility of a superconducting solution for the magnet system. This paper focuses on the analysis of the TF magnets thermal behavior. In particular, utilizing only the room available in the resistive design and referring to one of the most severe scenario envisaged for FAST, the minimum temperature margin in the coil has been calculated for a thermal load distribution on winding and cable jacket due to nuclear heating only.

[en] The recently developed 4C thermal-hydraulic code, currently under validation, is used to compute the temperature margin of the superconducting NbTi toroidal field coil of the ITER satellite tokamak JT-60SA, for the nominal burn operation of the machine. Repetitive conditions in the simulation are reached after two plasma pulses. For given (computed) thermal load distribution on winding and coil case due to nuclear heating only, helium temperature ∼4.4 K at the winding inlet, reference strand, and the most recent conductor and winding layout, the computed minimum margin occurs in pancake no. 7 at the peak field location and turns out to be ∼0.2-0.3 K above the design value of 1.2 K.

[en] The 18 D-shaped Nb-Ti toroidal field (TF) coils for the JT-60SA tokamak will each be 7 m high and 4.5 m wide. Together they will generate an on-axis field of 2.25 T. All the main contracts for their manufacture are now in place, with manufacturing split primarily between sites in Japan (superconducting strand), Italy (conductor cabling and jacketing, casings fabrication and coil winding, and integration), and France (support structures, coil winding and integration, and final coil cold testing). This paper will summarize the key aspects of the design of the coils and the current status of manufacture on each area of the manufacture of the TF coils. A simple overview of the overall schedule for their completion is included. (authors)

[en] The Divertor Tokamak Test (DTT) facility is an experimental facility under design and construction at ENEA C.R. Frascati. The aim of DTT is to investigate the power exhaust problem in a tokamak, providing possible alternative divertor solutions, with respect the conventional one, which can be extrapolated to DEMO fusion reactor. One of the main components of DTT facility is the Vacuum Vessel (VV), which has the function of providing an enclosed, vacuum environment for the plasma, acting also as a first confinement barrier. Starting from geometrical constraints, imposed by the desired plasma scenario and the configuration needed for the magnetic coils, the conceptual design of the VV was developed and dedicated research activities were carried out to verify design choices. A multiphysical approach was adopted with the aim of identifying a feasible and reliable solution for the VV, taking into account functional and design requirements, relevant aspects and issues. In particular, to assess and to optimize the VV design, fluid-dynamic, thermal and structural analyses were carried out, which are presented in the paper, as well as the progresses of the design.

[en] In the framework of the Broader Approach Agreement, Europe is involved in the design activities for the Japanese Tokamak JT-60SA, investigating, among several issues, the operation of the superconducting TF magnets and their subsystems, aimed at the reactor conceptual design definition. In particular, one of the main critical aspects to study is the heating of the conductor due to both direct component of energy deposited by neutrons and by secondary gamma generated during plasma operation, and heat generated by the radiation on casing and transferred to the winding pack. Indeed, the operating temperature and the relevant temperature margin (i.e. the operating safety margin) of the magnet will depend strongly on the heat loads and on the capability of the coolant to remove it. Furthermore, the heat power to the conductor will depend on several aspects, namely the thickness of insulating material, the mass flow rate of helium flowing in the conductors and its thermodynamic properties at operating conditions, and the layout of the superconductors constituting the winding. Moreover, a crucial aspect in the final design will be the presence and position of the casing cooling channels. In this paper a 2D sensitivity analysis of heat transfer from casing to winding pack with respect to cooling channels number and position is presented, based on the reference layout of the magnet. As a result, we evaluated the optimum number and positioning of cooling channels needed, as a trade-off between magnet operating limits and available cryogenic power and if, at limit, they could be even neglected in normal operation, keeping dwell-time within reasonable values.

[en] In the present study a complete neutronic analysis has been performed for the current design of the JT-60SA toroidal field coil (TFC) system. The MCNP5 Monte Carlo code has been used to calculate the nuclear heating, neutron spectra and absorbed dose in the TFC components, assuming a DD neutron emission rate of 1.5 x 10¹⁷ n/s (and 1% DT). Nuclear heating of the winding pack is lower than 0.3 mW/cm³ and the maximum nuclear heating of the TFC case is 0.4 mW/cm³. The overall nuclear heating, including the safety margin, is less than 8 kW. Spatial distribution of the nuclear heating has been provided along poloidal, radial and toroidal directions as to be used for thermo-hydraulic analysis and the design of TFC system. The absorbed dose to insulator is as low as to avoid the replacement during the whole life of the machine. Neutron fluxes have been used as input for a preliminary activation analysis performed with FISPACT inventory code. Activity and contact dose rates have been calculated at different cooling times, after 10 years of operations in some representative zone of the winding pack and the case. All the TFC materials can be easily recycled within the first day after shutdown and the hands-on recycling is possible within less than 30 years.