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Study on Heat Transfer Inside a Blocked Region of a Rod Bundle During a LOCA - Model and Experimental Approach

Abstract : 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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https://hal.univ-lorraine.fr/hal-03530147
Contributor : Michel Gradeck Connect in order to contact the contributor
Submitted on : Monday, January 17, 2022 - 5:17:15 PM
Last modification on : Wednesday, January 19, 2022 - 3:36:31 AM
Long-term archiving on: : Monday, April 18, 2022 - 9:10:53 PM

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Juan-David Peña-Carrillo, Tony Glantz, Georges Repetto, Alexandre Labergue, Michel Gradeck. Study on Heat Transfer Inside a Blocked Region of a Rod Bundle During a LOCA - Model and Experimental Approach. 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics NURETH-17, Sep 2017, Xian, China. ⟨hal-03530147⟩

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