Study on Heat Transfer Inside a Blocked Region of a Rod Bundle During a LOCA - Model and Experimental Approach
Résumé
This study focuses on the coolability of a partially deformed fuel assembly of a PWR reactor through an accident causing the loss of coolant primary (LOCA). Beyond the dryout point during the reflooding phase, the core is cooled mainly by the flow of superheated vapor and dispersed droplets. The thermal-hydraulic features of this two-phase flow, as well as the damaged cladding characteristics (blockage ratio, blockage length), are important parameters to estimate the heat exchanges and nuclear core coolability through the fuel assemblies.
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