The Fusion Advanced Study Torus (FAST), with its compact Tokamak design, high toroidal field and plasma current, will face many of the problems that ITER will meet and will anticipate many DEMO relevant physics and technology issues. The Design Upgrade of the Vessel and In-Vessel Components is presented in this paper. Relevant modifications were performed to the Vacuum Vessel (VV) and to the Plasma Facing Components (PFCs), i.e. the First Wall (FW) and the Divertor. The VV was modified to insert active reduction coils (ARC), between VV and the toroidal field (TF) coils to keep toroidal field magnet ripple lower than 0.3% and to allow Remote Handling for the FW and the Divertor. The FW, was modified to house coils for ELMs control and other plasma instabilities. A 3D thermo-hydraulic analysis using ANSYS code was performed to check FW heat removal capability. A new Divertor was designed to withstand the largest thermal loads of the high performance, low density, H-mode and to be able to comply with a recent magnetic topology called as "Snow Flake", increasing up a factor 4 the flux expansion. An exhaustive 3D thermo-hydraulic analysis using ANSYS code was carried out to show the capability of the Divertor to comply these high requirements. Design criteria were satisfied by present components of the upgraded machine. © 2013 Elsevier B.V.

Vessel and In-Vessel Components Design Upgrade of the FAST machine

Ramogida, G.;Maddaluno, G.;Cucchiaro, A.;Crisanti, F.;Brolatti, G.;Roccella, S.;Crescenzi, F.
2013

Abstract

The Fusion Advanced Study Torus (FAST), with its compact Tokamak design, high toroidal field and plasma current, will face many of the problems that ITER will meet and will anticipate many DEMO relevant physics and technology issues. The Design Upgrade of the Vessel and In-Vessel Components is presented in this paper. Relevant modifications were performed to the Vacuum Vessel (VV) and to the Plasma Facing Components (PFCs), i.e. the First Wall (FW) and the Divertor. The VV was modified to insert active reduction coils (ARC), between VV and the toroidal field (TF) coils to keep toroidal field magnet ripple lower than 0.3% and to allow Remote Handling for the FW and the Divertor. The FW, was modified to house coils for ELMs control and other plasma instabilities. A 3D thermo-hydraulic analysis using ANSYS code was performed to check FW heat removal capability. A new Divertor was designed to withstand the largest thermal loads of the high performance, low density, H-mode and to be able to comply with a recent magnetic topology called as "Snow Flake", increasing up a factor 4 the flux expansion. An exhaustive 3D thermo-hydraulic analysis using ANSYS code was carried out to show the capability of the Divertor to comply these high requirements. Design criteria were satisfied by present components of the upgraded machine. © 2013 Elsevier B.V.
PFC;FAST;ANSYS code
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/5005
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