▶ Date: 2022. 11. 24(THU)

▶ Time: 4:30 ~ 5:30PM

▶ Venue : Research Building I, #310 / Offline

▶ Speaker: Dr. Seok Kim(KAERI)

▶ Title: Measurement Techniques for Local Thermal-hydraulic Phenomena in Nuclear Reactor Safety System

▶ Abstract:

A multi-dimensional behavior of a two-phase flow is important for the thermal hydraulic safety of a nuclear reactor and for the development of a system analysis code and a local CFD code. Large pools of water near atmospheric pressure have been incorporated into several advanced nuclear reactor designs. These pools provide a heat sink to remove heat from the reactor or containment by natural circulation and are a source of water for core cooling. Thermal stratification is formed in horizontal fluid layers with different temperatures, where the warmer fluid layers are situated above the cooler fluid layers. The key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The current experimental data can be directly used to develop and validate turbulence models in a pool boiling. Furthermore, the present experimental results can help characterize the performance of boiling models on heated surfaces and provide benchmark data to validate the calculation performance of a component scale analysis code and CFD code.