Abstract
An accurate prediction of the interphase behaviors of the vertical gas-liquid subcooled boiling flow is meaningful for the first loop of a nuclear power plant (NPP). Therefore, the interphase behaviors including the bubble size distribution in the first loop of the NPP are analyzed, evaluated, and validated using various wall boiling models coupled with the population balance model (PBM) kernels in this paper. Firstly, nondimensional numbers of the first loop of the NPP and DEBORA (Development of Borehole Seals for High-Level Radioactive Waste) experiment test cases are analyzed with approximation. Secondly, five active nucleation site density models Nn coupled with the PBM kernel combination, four kernel combinations (C1~C4) with the Nn models are calculated and analyzed. Lastly, various behaviors including the bubble size distribution Sauter mean diameter (SMD) dp, void fraction α, gas superficial velocity jg, and liquid superficial velocity jl are compared and validated with the experimental data of the DEBORA-1 (P = 2.62 MPa). The results indicate that the two Nn models are suitable for the calculations of thefirst loop of the nuclear power plant. For instance, for the bubble size distribution SMD dp, the specified Nn model with C1 (maximum relative error 9.63%) has relatively better behaviors for the first loop of the NPP. Especially, the combination C1 is applicable for the calculation of the bubble size distribution dp, void fraction α and liquid superficial velocity jl while C4 is suitable for the calculation of the gas superficial velocity jg. These results can provide guidance for the numerical computation of the subcooled boiling flow in the first loop of the NPP.
Highlights
The gas-liquid subcooled boiling could occur in the first loop of the nuclear power plant (NPP) which highly affects the reactor safety and efficiency [5]
The population balance model (PBM) coupled with the EE two-fluid approach makes a tradeoff between the EE approach and EL method [12,13]
It is necessary to analyze, evaluate and validate the wall boiling models coupled with PBM kernels in the vertical gas-liquid subcooled boiling flow for the first loop of the NPP
Summary
Gas-liquid subcooled boiling flow is of enormous interest in many industrial applications like the thermal engineering systems, electronic systems, chemical reactors, and nuclear reactors [1]. An accurate prediction on the gas-liquid subcooled boiling flow is crucial for the operation safety and efficiency of the specific high-pressure application [4]. The gas-liquid subcooled boiling could occur in the first loop of the nuclear power plant (NPP) which highly affects the reactor safety and efficiency [5]. The phenomenon of the departure from nucleate boiling (DNB) relates with the critical heat flux (CHF), and the validations of the improvement on the CHF in the design of the nuclear power plant are expensive and time-consuming [6]. The computational fluid dynamics (CFD) approaches in the gas-liquid flow gradually gains popularity due to the development of computer technologies [7]
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