post critical heat transfer and fuel cladding oxidation

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ID: 255689
2016
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Abstract
The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.
Reference Key
caha2016actapost Use this key to autocite in the manuscript while using SciMatic Manuscript Manager or Thesis Manager
Authors ;Vojtěch Caha;Jakub Krejčí
Journal BMC Bioinformatics
Year 2016
DOI
10.14311/AP.2016.4.0008
URL
https://ojs.cvut.cz/ojs/index.php/APP/article/view/3723
Keywords
engineering (general) civil engineering (general)

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