Abstract
The Divertor Tokamak Test (DTT) facility, a fully superconducting nuclear fusion reactor being built in Italy, will contribute to address the power exhaust problem in EU DEMO perspective. A lot of flexibility of operation will be demanded to the machine, which should be capable to tackle also severe transients such as plasma disruptions. In this work, the 4C thermal-hydraulic code is used to compute the temperature margin during a plasma disruption, using as input the heat generated into the Toroidal Field coil casing and transferred to the winding pack, and the possibility that this leads to a quench of the magnet is studied. The results of the analysis will give important feedbacks for the design of the quench protection system, e.g., avoiding to trigger a fast current discharge right after the disruption, as well as for the machine operation, e.g., assessing the required re-cooling time of the magnets after a disruption.
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