A CHF correlation and flow pattern observation on a downward-Facing boiling surface
Proceedings of The 20th Pacific Basin Nuclear Conference
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After the Fukushima Daiichi nuclear accident in 2011, the heat removal capability of the in-vessel retention external reactor vessel cooling (IVR-ERVC) process for the advanced light water reactors (ALWRs) becomes a popular topic of the nuclear safety analysis field. The IVR-ERVC is a strategy to mitigate the consequences of the severe nuclear accident if the reactor core melts and relocates to the lower head of the reactor pressure vessel (RPV). In order to understand the physical phenomenon at the outer cooling surface of the RPV, a new heat transfer experiment was designed and established to investigate the local phenomena of heat transfer, boiling, and critical heat flux (CHF) at the downward-facing surface. The test facility can simulate the partial heating by the high temperature debris of molten fuels, and the experiments were performed under pool and forced impinging flow conditions. Different injection distances from the coolant inlet to the heating surface (Pool boiling, 8, 12, 16, and 20 cm) were tested. Different inclining angles were also discussed in this study. With the test results, the shorter inlet distance performed the higher CHF limit. The larger inclining angle had performed higher CHF limit, because the larger inclining angle induced the bubbles departing faster and easier from the heating surface. A new CHF model with downward-facing inclined angles has been derived of this study.
IVR, Downward-facing, Boiling, CHF, Heat transfer
A CHF correlation and flow pattern observation on a downward-Facing boiling surface.
Proceedings of The 20th Pacific Basin Nuclear Conference, 1