Hyoung Kyu Cho

Liquid film thickness measurements under temperature-varying conditions in a two-phase flow are of great importance to refining our understanding of two-phase flows. In order to overcome the limitations of the conventional electrical means of measuring the thickness of a liquid film, this study proposes a three-electrode conductance method, with the device fabricated on a flexible printed circuit board (FPCB). The three-electrode conductance method offers the advantage of applicability under con

In the direct vessel injection (DVI) system downcomer, the direct emergency core coolant (ECC) bypass is activated during the reflood phase of a large-break loss-of-coolant accident (LBLOCA) by the interaction between the downward-flowing liquid-film and the transverse gas flow. Direct ECC bypass is reportedly the major bypass mechanism of ECC, and various experiments have been performed to obtain detailed information about the ECC bypass in a DVI downcomer. These lead to a proposed new scaling

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in ac

Unlike land-based nuclear power plants, a marine or floating reactor is affected by external forces due to ocean conditions. These external forces can cause additional accelerations and affect each system and equipment of the marine reactor. Therefore, in designing a marine reactor and evaluating its performance and stability, a thermal hydraulic safety analysis code is necessary to consider the thermal hydrodynamic effects of ship motion. MARS, which is a reactor system analysis code, includes

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal-hydraulic code, named CUPID, has been developed. In the present study, the CUPID code was applied for the simulation of the PASCAL test facility constructed with an aim of validating the cooling and operational performance of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor + (APR+), which is intended to compl

An experimental work was conducted in a reduced-scale downcomer annulus of a nuclear reactor pressure vessel in order to investigate the behavior of liquid films under emergency core coolant (ECC) bypass conditions and to obtain high definition data for validating two-phase flow prediction capability of computational fluid dynamics (CFD) codes. Air and water at ambient pressure and temperature were used as a working fluid, creating a multidimensional two-phase flow in the test section. The main