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Conference Papers Year : 2017

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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Dates and versions

hal-03530147 , version 1 (17-01-2022)

Identifiers

  • HAL Id : hal-03530147 , version 1

Cite

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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