Critical heat flux for gap boiling on the outer surface of a fully submerged reactor vessel

摘要

The crical heat flux (CHF) for gap boiling in a verical annular channel that formed between a simulated reactor vessel and a scaled thermal insulation structure was studied experimentally to obtain basic information for assessing the thermal margins for in-vessel retention under severe accident conditions in an advanced nuclear reactor. Correlation euations for the spatial variation of the crtical heat flux along the external surface of the test vessel were derived from the CHF data measureed under both saturated and subcooled bioling conditions. The critical heat flux was found to have a nearly constant value in the bottom center region of the vessel. It decreased to a minimum value near the minimum gap of the cheannel and then increased monotonically toward the equator of the vessel. Results of this study were applied along with the calculated themal load to assess the thermal margins for external cooling of an advanced light water reactor in a postualated core-meltodown accient.