Keynotes and Plenaries

Keynote Speaker: Prof. Jiyun Zhao,
Department of Mechanical Engineering,
City University of Hong Kong, Hong Kong, China
Dr. Jiyun Zhao is an associate professor in Department of mechanical engineering at City University of Hong Kong. He received his PhD degree in nuclear science and engineering from MIT in 2005, and his master degree from Tsinghua University in 1999. He was a senior engineer working at Framatome (AREVA NP) North America Headquarters located in Lynchburg, Virginia from 2005 to 2010. He was appointed as an assistant professor in Nanyang Technological University, Singapore from 2010 to 2014. His research interests are nuclear thermal hydraulics and safety. He had published more than 120 papers in the leading journals and conferences. He received best paper awards from Progress in Nuclear Energy, Applied Energy etc. He is serving as an advisory editor for Annals of Nuclear Energy, and associate editor for Heat Transfer Engineering (Taylor & Francis, Inc.).
Keynote Title: Critical heat flux enhancement with micro/nano coated surfaces
Critical heat flux (CHF) is one of key concerns in water cooled reactors which often sets the design and operation limits of the reactors. It has been verified that the surface intrinsic wettability and surface structures have significant impact on the CHF and boiling heat transfer. Extensive researches have been conducted by different research groups around the world to improve the CHF through surface modifications. However, for those modified surfaces with hydrophilic nano/micro structures, the impact of surface wettability and structures on CHF are usually coupled together. To distinguish their individual impacts in order to better understand the underlying mechanism, it is worth conducting researches in decoupling them. In this research, surfaces coated with different micro/nano structures are compared through extensive experiments in order to identify the underlying mechanism of CHF enhancement for individual factor. In addition, although the surface orientation has very significant impact on the CHF, the research is still very limited and most of the research so far used smooth surface and the range of orientation angles is narrow. In this research, both bare substrates and the micro/nano coated surfaces are experimentally investigated with the orientation change from 0° to 180°. Finally, to comprehensively understand the underlying mechanism of CHF, molecular dynamics (MD) simulations that study the boiling process from the nanoscale point of view are carried out. Some atomic level information about CHF enhancement mechanism, which are unavailable using conventional method, is obtained.
Keynote Speaker: Dr. Dirk Lucas,
Institute of Fluid Dynamics,
Helmholtz-Zentrum Dresden – Rossendorf (HZDR), Dresden, Germany
Dr. Dirk Lucas is head of the Computational Fluid Dynamics (CFD) division at the Institute of Fluid Dynamics of the Helmholtz – Zentrum Dresden-Rossendorf (HZDR), Germany. His research focuses on the development and validation of CFD-models for multiphase flows in medium and large-scale applications, e.g. in chemical engineering and nuclear reactor safety. In particular, he is interested in fundamental phenomena in poly-disperse bubbly flows and in a corresponding modelling in the frame of CFD codes.
Dirk Lucas studied physics at the Technical University of Dresden from 1983 to 1988 and continued with his doctoral thesis at the Technical University of Zittau. He received his PhD in 1991 from the Technical University of Dresden. From 1992 to 2011 he worked as a research fellow at the Research Center Dresden-Rossendorf (now Helmholtz-Zentrum Dresden – Rossendorf) and became head of the CFD department in 2012. He is vice chair of the Virtual International Research Institute of Two-Phase Flow and Heat Transfer (VIR2AL: and member of the editorial boards of the journals “Multiphase Science and Technology” and “Experimental and Computational Multiphase Flow”. He is the main organizer of the annual Multiphase Flow Conference and Short Course. Dirk Lucas is author or co-author of 185 scientific papers in peer-reviewed journals.
Keynote Title: Multiphase CFD for Nuclear Safety Research
While according to the state of the art safety analyses related to nuclear reactor thermal hydraulics are done using system codes, CFD becomes more and more important as a tool to support such analyses. Often multiphase flows are involved in the flow situations that have to be considered. For medium and large scales, as they are typical in nuclear reactor safety research, the Euler-Euler (E-E) approach is most suited and most frequently used. However, E-E-CFD is not yet mature. A consolidation of closure models is required to enable reliable predictions. The baseline model concept is a promising way to reach such a consolidation. A baseline model for poly-disperse bubbly flows is established and implemented in the OpenSource CFD package OpenFOAM (Foundation release). The sustainable development requires quality insurance and continuous maintenance of the corresponding OpenFOAM-add-ons which is enabled in the GitLab environement. An automated workflow helps efficiently to support the step-by-step improvement of the baseline model. The lecture presents the baseline model concept and the general strategy for its further development. This is illustrated by the example of the baseline model for poly-disperse flows. Advantages as well as challenges related to the use of Open Source software in NRS are discussed.
Keynote Speaker: Dr. Tomoyuki Sugiyama,
Nuclear Safety Research Center,
Japan Atomic Energy Agency (JAEA), Japan
Dr. Tomoyuki Sugiyama joined the Japan Atomic Energy Research Institute (currently Japan Atomic Energy Agency; JAEA) in 1996 after earning his PhD in mechanical engineering from Tokyo Institute of Technology. He was engaged in nuclear fuel safety research and his primary work was investigation of transient fuel behavior using the research reactor NSRR and hot laboratories in JAEA.
He had been dispatched to the OECD Halden Reactor Project in Norway from 2001 to 2002 to join the international fuel and materials irradiation program. He contributed to data extension on high burnup LWR fuel behavior under normal operation and transient conditions.
In 2012, he joined the discussion for new regulatory requirements for Japanese power reactors conducted by the Nuclear Regulation Authority (NRA) of Japan established in 2012 after the Fukushima Daiichi NPS accident. He also served for the safety review of nuclear power stations as a regulatory specialist of the NRA for 2 years from 2014. He mainly contributed to the review of effectiveness evaluation of severe accident measures newly equipped at power stations in accordance with the new regulations.
His current position is the head of the Reactor Safety Research Division in the Nuclear Safety Research Center of JAEA. He is also leading the Severe Accident Research group.
Keynote Title: Expansion of safety research scope at the Nuclear Safety Research Center of JAEA based on the lessons learned from the Fukushima Daiichi NPS accident
The Nuclear Safety Research Center of Japan Atomic Energy Agency (JAEA-NSRC) technically supports nuclear regulation. Our research activities cover extensive areas including reactor, materials, fuel cycle, environment, waste. This presentation looks back on the impact of the Fukushima Daiichi NPS (1F) accident on our activities specifically on reactor safety research.
Before the 1F-accident, Japanese regulatory requirement for reactor facilities was safety ensuring against Design Basis Accidents and, accordingly, our primary research target was the prevention of Severe Accident (SA). The 1F-accident, however, revealed our wrong awareness of SA being a hypothetical event and triggered expansion of our research scope. Now we put much resource into the research for SA mitigation, source term evaluation, off-site consequence evaluation and accident preparedness. In addition, JAEA is serving as a channel to share 1F-accident information internationally by operating OECD/NEA joint projects aiming at 1F information collection, analysis, evaluation and so on.
What important in future is maintaining the adequate level of human and financial resources for safety research, including that for SA, and keeping motivation to improve safety. JAEA-NSRC will utilize our experimental facilities, including new ones built after the 1F-accident, for safety research of existing nuclear installations as well as of newly developed technologies, such as accident tolerant fuels and small modular reactors, and educate researchers who can tackle specific phenomena and can overview the safety of nuclear systems.
Keynote Speaker: Prof. Shun-Chi Wu,
Department of Engineering and System Science,
National Tsing Hua University, Taiwan
Prof. Shun-Chi Wu received the B.S. and M.S. degrees in Engineering and System Science from National Tsing Hua University, Hsinchu, Taiwan, in 2000 and 2002, respectively, and the Ph.D. degree in Electrical Engineering and Computer Science from University of California, Irvine, in 2012. From 2003 to 2007, he was a research assistant at National Space Organization, Hsinchu, Taiwan. In 2013, he was employed at IMEC, Taiwan Co., of Hsinchu, Taiwan, where he was involved in designing algorithms and architectures for several wearable devices. He is currently an associate professor of Engineering and System Science at National Tsing Hua University. His research interests lie in applying signal/image processing and machine/deep learning techniques to analyze power plant signals to enable various operational needs. He received the MOST Future Technology Award in 2020 and the Outstanding Teaching Award of NTHU in 2021.
Keynote Title: Initiating event detection and identification in nuclear power plants with data-driven techniques
An initiating event (IE) is an event that creates a disturbance in a nuclear power plant (NPP) and can potentially lead to core damage. When an IE occurs, mitigation actions should be taken in time to prevent its escalation into a severe accident. Due to the situation’s urgency, the actions taken in general are intended to soothe the “symptoms” (i.e., the abnormal tendencies shown on the primary plant sensors) incurred by the IE without actually knowing it. However, to successfully bring the plant back into safe shutdown conditions, its identification is required. This task needs plant operators to thoroughly examine tendencies in various plant sensors’ readings, but stressful and chaotic circumstances make this manual examination challenging. In the past years, we have focused on applying data-driven methods to realize timely detection and identification of IEs. Several approaches based on machine learning or deep learning techniques have been proposed and presented. In this talk, the application prerequisites and the pros and cons of these approaches will be thoroughly discussed. Some challenges to propel their applicability in practice are also covered.
Keynote Speaker: Prof. Dong Yujie,
Institute of Nuclear and New Energy Technology,
Tsinghua University, China
Dr. Dong is a Professor in nuclear engineering at the Tsinghua University, Beijing, China, where he earned his Bachelor’s degree in fluid mechanics and PhD in nuclear engineering. Since 1997, he has worked to develop advanced nuclear reactors at the Institute of Nuclear and New Energy Technology (INET), Tsinghua University. He has served as the head of the Division of Reactor Thermal-Hydraulic, and head of the Division of Reactor Physics, Thermal-Hydraulics and System Simulation. Since 2006, He has been focusing on research and development of the high temperature gas-cooled reactor (HTGR) technology all the time. Currently, he is the Deputy Director and Deputy Chief Engineer of INET in charge of HTGR projects. He has deeply involved in the entire process of planning and implementation of the HTR-PM demonstration power plant project, including concept development, engineering design, component tests, construction, commissioning, etc.
Keynote Title: R&D of the HTR-PM and construction of the demonstration power plant
The HTR-PM is a modular high temperature gas-cooled reactor with a pebble-bed core. Helium is used as the coolant and each fuel element is composed of a large number of TRISO coated fuel particles which is high temperature resistant. A simple cylindrical core is chosen. A steam turbine cycle is selected for electricity generation or cogeneration. The temperature of hot helium is 750 ℃ and that of fresh steam is 570 ℃. The thermal power of each module is 250 MW. The fresh fuel elements are loaded and the spent ones are unloaded continuously every day. This reactor possesses inherent safety. It can be shutdown automatically due to negative feedback caused by temperature increase. Its decay heat can be removed in natural ways, including heat radiation, conduction and convection. The fission products are contained within the coated particles. It can extend the application of nuclear energy,such as cogeneration, heat utilization for various industrial processes. A demonstration power plant consisting of two modules and one steam turbine-generator has been built and the comprehensive commissioning is undergoing. The R&D and progress of the demonstration plant are presented.