Abstract
A Tokamak type fusion reactor is characterized by very high heat loads to plasma facing surfaces of components surrounding the plasma. In order to maintain structural integrity of in-vessel components, it is essential to provide sufficient cooling of those components during normal operation as well as during accident conditions. In relation to the ITER-FEAT design efforts, component thermal responses have been evaluated for a number of specified transient events. In this paper results are presented from thermal-hydraulic analyses of two postulated transients in the divertor cooling loop system, (i) trip of the main coolant pump in the divertor primary heat transport system, and (ii) loss of heat sink in the divertor cooling system; trip of the secondary side pump. The analyses have been made using the ATHENA thermal-hydraulic system computer code.
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