Abstract
For a water cooled reactor, the key thermal-hydraulic parameters span a wide range corresponding to different CHF regimes. Under accident conditions, due to the flow regime transition and interchannel mixing effect, the corresponding CHF can transition from the DO to DNB regime. In order to continuously and accurately predict DNB and DO regime CHF under wide parameter range for rod bundle channel, a comprehensive CHF mechanistic model covering the DNB and DO regime CHF prediction is established based on the rod bundle CHF-regime criterion. The DNB regime CHF mechanistic model of superheated liquid layer depletion under turbulence fluctuation bubbles and the mature DO regime CHF mechanistic model are combined to form the comprehensive CHF model. Furthermore, the comprehensive CHF model is assessed by 5 × 5 rod bundle CHF experimental data independently obtained by the Nuclear Power Institute of China (NPIC). The statistical evaluation and parametric trend analysis show that the maximum mean error of P/M is within ±22%, and the local pressure, mass flux, and quality do not have any effects on the average deviations of the predicted flux P from the measured flux M. This indicates that the comprehensive CHF mechanistic model can accurately and continuously predict the DNB and DO regime CHF in the rod bundle channel.
Highlights
Critical heat flux (CHF) is an important thermal safety limit in the research and development of nuclear fuel assemblies and reactor thermal-hydraulic design and safety analysis
According to the different flow regimes and heat transfer characteristics corresponding to CHF, the flow boiling crisis in rod bundle channel can be divided into departure from nucleate boiling (DNB) regime and dryout (DO) regime (Tong, 1967)
It can be seen that the comprehensive CHF mechanistic model developed in this study can accurately and continuously predict the DNB and DO regime CHF in the rod bundle channel
Summary
Critical heat flux (CHF) is an important thermal safety limit in the research and development of nuclear fuel assemblies and reactor thermal-hydraulic design and safety analysis. According to the different flow regimes and heat transfer characteristics corresponding to CHF, the flow boiling crisis in rod bundle channel can be divided into DNB regime and dryout (DO) regime (Tong, 1967). For the pressurized water reactor (PWR), subcooled nucleate boiling occurs at the hot channel exit under normal operation conditions. The DNB regime CHF is the most likely to occur due to the low vapor quality in the channel. There are several kinds of flow regimes existing in the rod bundle channel. Due to the interchannel mixing mechanisms (Xiong et al, 2020) and the cross flow caused by the mixing vanes between adjacent open channels (Qu et al, 2019), the flow regime
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