Film boiling is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In this study, an experimental test facility was designed and constructed with the purpose of performing high-temperature transient pool boiling quenching experiments for stainless steel (SS) and zirconium (Zr) cylindrical test samples. The fabricated samples have three embedded thermocouples between the inner surface of the rods and the boron nitride (BN) filler material. The focus of this research is to study the effects of liquid subcooling, surface oxidation and surface materials for the prescribed materials on the minimum film boiling temperature, Tmin. The samples were heated to a temperature well above the Tmin (up to 550oC) then plunged vertically in a quiescent pool of subcooled and saturated distilled water. A data acquisition system was used to record the temperature of the embedded thermocouple locations over time. Data reduction was performed by using an inverse heat conduction code (DATARH) to calculate the surface temperature and the corresponding wall heat flux. A visualization study using a high-speed camera was conducted. Additionally, a characterization study using X-ray Diffraction (XRD), Scanning Electron Microscopy (SEM), and Energy Dispersive X-ray Spectroscopy (EDS) methods was performed to quantify surface conditions. Results indicate that liquid subcooling has a strong influence on Tmin. The visualization study shows a very thin vapor formation around the rod for higher subcooling cases that explains the enhancement in the heat transfer. Also, the surface characterization shows the difference in surface condition of the SS and Zr that leads the vapor bubbles to depart differently in the nucleate boiling regime. Furthermore, the effect of surface oxidation is noticeable in the Zr rod compared to SS rod due to the oxidation kinematic of the substrate material. Finally, it is found that the surface properties have a significant impact on Tmin.
|Published - 2017
|17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China
Duration: Sep 3 2017 → Sep 8 2017
|17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
|9/3/17 → 9/8/17
All Science Journal Classification (ASJC) codes
- Nuclear Energy and Engineering