RESPONSE OF HEATING ROD IN WATER POOL FOR PREDICTING CRITICAL HEAT FLUX (CHF)

摘要

Critical heat flux is paramount from consideration of reactor core thermal hydraulics and ensuring fuel pin cladding integrity. Conventionally, CHF measurements are carried out in out of pile test setups using locally placed thermocouples and also by indirect measurement of change in electrical resistance. These measurements have limitations which are well documented in literature. The present knowledge base of acoustic/vibration response in the core due to CHF is also considered as alternate methodology to detect the onset of departure from nucleate boiling (DNB) and in determination of CHF. Pool boiling experiments were performed to create condition for critical heat flux to occur in the pool and to study fluid structure interactions during different phases of boiling.