Experimental investigation of flow boiling CHF of heater rod with axially cosine-shaped power distribution under heaving conditions

Floating Nuclear Power Plants (FNPPs) have been attracting attention as a means of supplying power to hard-to-reach areas or for various purposes. Since FNPPs are located in the ocean environment, they are subjected to rotational and translational motion, resulting in thermal-hydraulic phenomena different from those of land-based reactors. Among these phenomena, it is crucial to investigate how the Critical Heat Flux (CHF) is affected under ocean conditions, as it is directly related to the safety margin of the reactor. Additionally, since the power distribution of nuclear fuel rod is non-uniform, experiments that consider this power distribution are necessary, but scarce. Therefore, in this study, a single heater rod with cosine-shaped power distribution was used to conduct experiments under vertical static and heaving conditions. High-pressure experiments using simulant fluid R134a were conducted to consider the Pressurized Water Reactor (PWR) operating conditions. The parametric effect of thermal-hydraulic parameters under the vertical static condition, and the effect of heaving on CHF were analyzed. A comparison with previous experimental results using heater rod with uniform power distribution in the same apparatus was made to analyze the effect of power distribution on CHF. The results are expected to be helpful for the design of FNPPs.