Critical heat flux (CHF) has been widely studied in the past decades because of its importance for nuclear power plant design. But most of the studies are based on flow under normal operating conditions for light water reactors. CHF under low flow and low pressure is of significance when considering operating transients and accidents. In this study, experimental study has been carried out on CHF for low flow rate and low pressure water flow in vertical bilaterally heated annuli. Parameter trends on CHF is discussed and a new predictive correlation was fitted based on the CHF data points. This study is meaningful for concerned nuclear engineering and similar experiment design.

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