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The design of RFX-mod2 experiment implies a major change and reconfiguration of the internal components of the present machine assembly. Critical aspects of the design regard the technical solutions needed to fulfil geometrical, vacuum and electrical requirements imposed by interface with existing components and by specific operating conditions. The paper presents an overview of the design choices and the proposed implementations, assessed by means of engineering analyses and experimental tests performed on mock-ups of the new components.

A major modification of the RFX-mod toroidal load assembly has been decided in order to improve passive MHD control and to minimise the braking torque on the plasma, thus extending the operational space in both RFP and Tokamak configurations. With the removal of the vacuum vessel, the support structure will be modified in order to obtain a new vacuum-tight chamber and the first wall tiles will be directly in front of the passive stabilising shell inside of it, so increasing both the poloidal cross section (by 28 mm radially) and the plasma-shell proximity. This implies the design of a new vacuum fit electromagnetic measurement system. The spatial resolution is constrained by the number of graphite tiles, 28 (poloidal) x 72 (toroidal), which cover the inner surface of the copper shell. The new local probes (of size 42 x 37 x 7 mm) will be installed in vacuum onto the copper shell, behind the graphite tiles, and shall operate up to a maximum temperature of 180°C to allow for baking cycles for first wall conditioning. Because of the reduced room available, triaxial pickup probes have been designed, with the additional advantage of allowing the minimization of alignment errors. The paper describes the detailed design of the new probe set, in particular highlighting the technological issues complying with different functional requirements: the thermo-mechanical interface with the copper shell, their protection, the electrical connection for the transmission of the signal with a minimum level of noise and the choice of materials suitable to the vacuum environment at relatively high temperature. The extensive tests carried out on probe prototypes to characterise their electromagnetic, thermal and vacuum behaviour are also reported.

This paper describes the final design of the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE) for the SPIDER experiment (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma) under construction at the Consorzio RFX premises. The STRIKE diagnostic will be used to characterise the SPIDER beam during short pulse operation (several seconds) to verify the degree of attainment of ITER requirements about the maximum allowed beam non-uniformity. After a preliminary design developed in the last few years, the complete STRIKE diagnostic system has been recently subjected to a final review and is now ready for construction. The main components of the system are: an array of 16 tiles made of Carbon Fibre Composite, which will be exposed to the high power density beam (of the order of 20 MW/m22) exiting from the SPIDER Beam Source; a supporting and positioning system, based on a set of structures made of stainless steel, with controlled moving systems, to be installed within the SPIDER vacuum vessel; a set of thermal, electric and thermo-graphic sensors to properly detect the operating conditions. The paper will focus in particular on the development of the engineering design of the supporting and positioning system, with a description of the relevant CAD and FEM analyses, and will give an overview of the complete system, with reference to the manufacture of CFC tile prototypes and to the choice of the complete set of sensors.

After 10 years of operation since its major modification, an upgrade of the RFX-mod experiment is presently under design. The scientific objective is the improvement of 3D physics studies through a more robust transition to higher confinement regimes in both Reversed Field Pinch (RFP) and Tokamak configuration obtained thanks to an advanced system for the active control of MHD instabilities. The main design driver requirements for this machine upgrade are the removal of the present resistive vacuum vessel and the enhancement of the ‘shell-plasma proximity’, to reduce the deformation of the last close magnetic surface and to improve the self-organized helical plasma regimes. The fulfillment of these requirements implies a major change of the internal components of the machine such as the replacement of the whole first wall, the change of the support system of the stabilizing shell and the modification of the present toroidal support structure to provide the function of vacuum barrier. In combination, other components of the machine will be upgraded, such as magnets and power supply, diagnostic systems and a NBI will be integrated. The paper presents an overview of the engineering design of the new components and highlights the critical aspects of the new torus assembly.

This paper is focused on the assessment of the thermal behaviour in nominal operating conditions of two components of the ITER-NBI ion source: the arc-discharge chamber and the filament cassette assembly. The study is based on hydraulic, thermo-mechanical and thermo-electrical calculations performed by means of 2D and 3D finite element (FE) models, with inputs coming partly from the ITER reference design documentation and partly from the design review activities presently in progress. The paper describes the results of the analyses, comparing the alternative cooling design with the reference one, and highlights some open issues that should be worked out before finalizing the design.

This paper presents the status of the design and manufacture of the main in-vessel subsystems of the magnetic diagnostics enhancement project for JET. The subsystems are conceived as several sets of pick-up coils supported by mechanical structures attached to existing components of the first wall, to be installed by remote handling. The design constraints and fundamental choices at the base of the project are illustrated. The technological solutions proposed for the construction, presently in progress, are described.

Recent studies on RFP and Tokamak devices call for an active control of the MHD and resistive wall modes to induce plasma mode rotation and to prevent mode phase locking. The results obtained on RFX, where slow rotation of phase locked modes has been induced, support the possibility of extending active MHD mode control through a substantial modification of the device. A new first wall with an integrated system of electric and magnetic transducers has been realised. A close fitting 3 mm thick Cu shell replaces the 65 mm Al shell. A toroidal support structure (TSS) made of stainless steel replaces the shell in supporting all the forces acting on the torus. A system of 192 saddle coils is provided to actively control the MHD modes. This system completely surrounds the toroidal surface and allows the generation of harmonic fields with m=0 and m=1 poloidal wave number and with a toroidal spectrum up to n=24.