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.
Highlights
The post critical heat flux heat transfer is encountered when the surface temperature becomes too high to maintain a continuous liquid contact, and the surface becomes covered by a continuous vapour blanket
It results in coverage of the heated surface by a continuous vapour film in the case of film boiling regime, or an intermittent vapour film in the case of transition boiling regime
The boundary between these post dryout heat transfer boiling regimes is the minimum of film boiling temperature, or Leidenfrost temperature
Summary
The post critical heat flux heat transfer is encountered when the surface temperature becomes too high to maintain a continuous liquid contact, and the surface becomes covered by a continuous vapour blanket It results in coverage of the heated surface by a continuous vapour film in the case of film boiling regime, or an intermittent vapour film in the case of transition boiling regime. Inverted annular film boiling is characterized by a vapour layer separating the continuous liquid core from the heated surface and usually encountered at void fractions below 40 %. Discrete liquid drops entrained in a continuous vapour flow; normally encountered at void fractions above 80 % The transition between these two cases is the slug flow film boiling. The nucleate boiling regime is immediately followed by film boiling regime in the nuclear reactor core [1]
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