Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Abstract

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the downcomer boiling experiment (DOBO) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change from a bubbly flow to churn and mist flows and a circulation of liquid accelerated by bubbles. The twophase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling.

Numerical simulation of air–water two–phase flow in vertical pipe using k-ε model

“…The pressure-gradient equation derived for single-phase flow can be modified for multiphase flow by considering the fluids to be a homogeneous mixture [20]. Thus, (1) (2) The pressure-drop component caused by friction losses requires evaluation of a two-phase friction factor. The pressure drop caused by elevation change depends on the density of the two-phase mixture which is usually calculated with Eq.…”

“…Two-phase flow in vertical pipe, are commonly used in many applications such as: petroleum industry, food processing, chemical industry, power plants like coal fired power plant and heat transferring of fluids. Two-phase flow and pressure distribution through the pipe have been studied numerically and experimentally by many researchers [1][2][3][4]. Cho et al [1] numerically analyzed the transient two-phase flow in nuclear reactor components, a three-dimensional thermal hydraulics code, named COPID.…”

“…Two-phase flow and pressure distribution through the pipe have been studied numerically and experimentally by many researchers [1][2][3][4]. Cho et al [1] numerically analyzed the transient two-phase flow in nuclear reactor components, a three-dimensional thermal hydraulics code, named COPID. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change from a bubbly flow to churn and mist flows and a circulation of liquid accelerated by bubbles.…”

Effects of surface orientation on nucleate boiling heat transfer in a pool of water under atmospheric pressure

“…Recently, the field of nuclear thermal hydraulics has begun to greatly benefit from full three dimensional computational fluid dynamics (CFD) for component level modeling Cho et al, 2012). It is obvious that for an accurate analysis of natural-circulation-driven thermal-hydraulic cooling systems with heat exchangers submerged in a large pool, the effects of surface orientation needs to be incorporated into the prediction model of nucleate boiling from a heated wall.…”

Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

“…It has been verified against standard conceptual problems of single-and two-phase flows and validated for thermal-hydraulic experiments in our previous studies [4][5][6]. The assessment strategy for the future verification and validation was outlined in Jeong et al [7].…”