Laishun Wang’s research while affiliated with Sun Yat-sen University and other places

The nucleation and growth of bubbles on homogeneous wetting surfaces have been extensively studied, but the intricate dynamics on hybrid wetting surfaces remain under-explored. This research aims to elucidate the impact of hybrid wettability on pool boiling heat transfer efficiency, specifically under downward-facing heating conditions. To this end, a series of hybrid wettability surfaces with varying hydrophilic and hydrophobic configurations are meticulously fabricated and analyzed. The study reveals distinctive interfacial phenomena occurring at the boundary between hydrophilic and hydrophobic regions during the boiling process. Experimental results indicate that surfaces with a higher proportion of hydrophilic to hydrophobic interfaces exhibit reduced superheat requirements and enhanced boiling heat transfer coefficients for equivalent heat flux densities. Furthermore, the rewetting characteristics of hybrid wettability surfaces are identified as pivotal factors in determining their critical heat flux (CHF). This investigation underscores the potential of hybrid wettability surfaces to optimize pool boiling heat transfer, offering valuable insights for the design and en-hancement of heat exchangers and other thermal management systems.

Phase change materials (PCMs) are promising in many fields related to energy utilization and thermal management. However, the low thermal conductivity and poor shape stability of PCMs restrict their direct thermal energy conversion and storage. The desired properties for PCMs are not only high thermal conductivity and excellent shape stability, but also high latent heat retention. In this study, the boron nitride nanosheets (BNNSs) were bridged by small amounts of GO nanosheets and successfully self-assembled into BNNS/rGO (BG) aerogels by hydrothermal and freeze-drying processes. The BG aerogels with interlaced macro-/micro-pores have been proven to be ideally suited as support frameworks for encapsulating polyethylene glycol (PEG). The obtained composite PCMs exhibit high thermal conductivity (up to 1.12 W m⁻¹ K⁻¹), excellent shape stability (maintain at 90 °C for 10 min), and high latent heat (187.2 J g⁻¹) with a retention of 97.3% of the pure PEG, presenting great potential applications in energy storage systems and thermal management of electronic devices.

Studies about using nanofluids to enhance the Critical Heat Flux (CHF) of In-vessel Retention (IVR) strategy in the third-generation reactor have been conducted extensively and show a significant CHF enhancement effect. However, low carbon steel SA508 used in the reactor vessel is easy to oxidize and the oxidation can lead to changes in the surface properties which may affect the CHF enhancement effect of nanofluids. In this study, pool boiling CHF experiments with low carbon steel SA508 surfaces were conducted in distilled water and nanofluids under different boiling time to investigate the CHF enhancement effect of nanofluids under low carbon steel surface oxidization condition. CHF in distilled water increases rapidly with boiling time due to the rapid surface oxidation during the boiling and the increase ratio can be nearly 2 due to the surface oxidation. CHF in nanofluids is stable and independent of boiling time. The difference between CHF in nanofluids and CHF in distilled water decrease to 17% under the longest boiling time conditions due to the surface oxidation. The deposition layer of nanoparticles on the surface leads to the capillary wicking and decrease in the nucleation site and thus, CHF is enhanced. This study is of great significance for exploring the actual effect of nanofluids on the CHF enhancement of IVR strategy.

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.