Kyushu University Academic Staff Educational and Research Activities Database
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Wei Liu Last modified date:2021.06.19

Associate Professor / 原子力エネルギーシステム
Department of Applied Quantum Physics and Nuclear Engineering
Faculty of Engineering

Graduate School
Undergraduate School

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Academic Degree
PhD in Engineering
Country of degree conferring institution (Overseas)
Field of Specialization
Thermal Hydraulics in Nuclear Reactor Systems
Total Priod of education and research career in the foreign country
Outline Activities
The situation surrounding nuclear energy by the influence of the Fukushima accident has reached a major turning point. Further safety security of nuclear power systems that using nuclear fission energy is required and the development of new advanced reactor systems is needed. In particular, in order to make the third-generation light-water reactors and future light water reactors being safer and more reliable, and in order to realize a high environmental compatible next-generation reactor system, it is necessary for us to evaluate aerosol transportation and to model the heat transfer and flow characteristics in multidimensional two-phase flow and to establish analysis methods in the thermal design of the reactor core and the safety evaluation at accidents over the entire system. For this reason, we conduct research, education, and social activities on "heat transfer and flow characteristics in gas-liquid two-phase flow including phase change", based on clarification of the basic physical mechanism. Recently, we have also conducted research on the utilization of waste heat.
Research Interests
  • Study on aerosol transportation behavior
    keyword : aerosol, transportation, mechanism, nuclear power plant
  • Development of a High-Temperature Air Generation System Using Evaporated Moisture from Low-Temperature Wastewater
    keyword : High temperature air generation, fehn, Low-Temperature Wastewater utilization
  • Development of simulation methods for boiling and critical heat flux
    keyword : boiling, critical heat flux, simulation
  • Development of liquid fuel assembly device to prevent core disruptive accidents in fast reactors
    keyword : fast reactors, core disruptive accidents prevention, liquid fuel assembly device
  • Experimental Study on the Mechanism of Forced Subcooled Flow Boiling Critical Heat Flux
    keyword : Subcooled Flow Boiling, Critical Heat Flux, Mechanism,Experimental Study
  • Research on flow and heat transfer characteristics in micro channels
    keyword : two phase flow, heat transfer, flow, boiling, pressure loss
Current and Past Project
  • As a core disruptive accident mitigation measure for large-sized sodium-cooled fast reactors, various passive safety systems have been proposed to mitigate the loss of reactor shutdown function events (ATWS). On the other hand, the concept of controlled material relocation (CMR), which has been adopted as a mitigation measure for core disruptive accidents in large fast reactors with a large fuel inventory, is a predesigned countermeasure to reduce the excess reactivity of the system and maintain the subcritical state by controlling the relocation of core materials during a core disruptive accident. The concept of controlled material relocation (CMR) is to incorporate the function to reduce the excess reactivity of the system by controlling the relocation of core materials in case of core disruptive accidents and to maintain the subcritical state as a design measure. In this study, based on the CMR concept, an assembly-type concept with liquid fuel enclosed in pins is proposed as a countermeasure to prevent core disruptive accidents in sodium-cooled fast reactors. In this study, we propose an assembly-type device with liquid fuel enclosed in pins as a preventive measure against core disruptive accidents (CDAs) in sodium-cooled fast reactors. It has a passive safety feature that makes the reactor subcritical before normal solid fuel damage. Furthermore, by using the system in combination with the existing severe accident prevention measures, it will contribute to the improvement of safety so that core damage can be regarded as an extremely unlikely event with a high level of confidence by thickening the independent protection lines with diversity and robustness.
  • It has been reported that the critical performance of boiling cooling is drastically reduced under irradiation simulating actual reactor conditions. If this happens in actual reactors, it will have a major impact on the variability of IVR technology (external cooling by submerging the entire reactor vessel in pool water), which is being developed as a safety measure in case of severe accidents such as a core meltdown. In this project, we will investigate the effect of irradiation on the reduction of the boiling cooling limit and develop a method to prevent the reduction of the cooling performance. Furthermore, by introducing a revolutionary honeycomb cooling method, we will develop a method that not only prevents the deterioration of the limit performance of IVR in supercritical water reactors but also dramatically improves it even under irradiation simulating the actual conditions.
Academic Activities
Membership in Academic Society
  • The Japan Society of Mechanical Engineering
  • Atomic Energy Society of Japan
  • The Heat Transfer Society of Japan
Educational Activities
I conduct lectures in undergraduate and graduate schools.
In addition to providing students with specialized knowledge through regular seminars, we give them problems to solve in order to promote retention of specialized knowledge and problem-solving skills.
Guidance is also provided for undergraduate thesis research, master's thesis research, and doctoral dissertation research. Students are divided into groups based on their research themes. Each group needs to report its progress every week.
Other Educational Activities
  • 2019.08.