Keynotes and Plenaries

Plenary I
Prof. Yassin A. Hassan
Distinguished Professor, Department of Nuclear Engineering, Texas A&M University, United States

Yassin Hassan is University Distinguished Professor, Regents Professor and the L.F. Peterson ’36 Chair II in Engineering. He is a professor in the Department of Nuclear Engineering and the J. Mike Walker ’66 Department of Mechanical Engineering, Texas A&M University. He is the Director of the Center for Advanced Small Modular and Microreactors (CASMR).
Prior to joining Texas A&M September 1986, he worked for seven years at Nuclear Power Division, Babcock & Wilcox Company, Lynchburg, Virginia. His research is in computational and experimental thermal hydraulics and advanced nuclear reactors.
He is a Member of the United States National Academy of Engineering and a Member of the Academy of Medicine, Engineering and Science of Texas and a fellow of the American Association for the Advancement of Science (AAAS), the American Nuclear Society (ANS) and the American Society of Mechanical Engineers (ASME). His awards include the 2022 ASME-Fluid Engineering Award, American Nuclear Society Seaborg Medal. He is a member of Slovenian Academy of Engineering.
Hassan was sworn in 2007 as a part-time technical judge to the Atomic Safety and Licensing Board Panel, US Nuclear Regulatory Commission. He is appointed a member of the Civil Nuclear Trade Advisory committee by Honorable Commerce Secretary Gina Raimondo, May 4, 2021. He is the recipient of the 2020 Grainger College of Engineering, University of Illinois Alumni Distinguished Award.
He received his master’s and doctorate in nuclear in nuclear engineering from the University of Illinois and a master’s in mechanical engineering from the University of Virginia.
Plenary Title: Advanced Techniques in Reactor Thermal Hydraulics and Safety
Recently, Computational Fluid Dynamics (CFD) simulations for the analyses of fluid flows are rapidly increasing with recent developments of supercomputers. To validate the computational results, high accuracy full-field experimental data is needed.

Flow field measurements in geometries that are typically found in nuclear engineering applications, such as wire-wrapped fuel assemblies, randomly packed beds, spacer-grids fuel assemblies and helical coil steam generators are complex. Performing flow measurements in such complex flow systems require a substantial effort to obtain a fundamental knowledge about the flow characteristics.

In this talk, several experimental techniques aimed at providing experimental databases with high quality, high spatial and temporal resolutions of interesting dynamics in nuclear application domains, will be discussed. The obtained high-fidelity experimental data will be used in the verification and validation (V&V) of computational fluid dynamics (CFD) codes. The modeling and simulation can be used for overseeing the continued safe and efficient operation of the current nuclear power plants and in establishing an effective licensing framework for a new generation of nuclear technologies.

Plenary II
Jason Chao
Chief Scientist and Vice President of Nuclear Resilience Group Inc., United States

Dr. Chao is the author of a book entitled “A Complete Perspective of Nuclear Energy” recently published by Nova Science. Currently, he is the Chief Scientist and Vice President of Nuclear Resilience Group Inc., responsible for the intellectual property of novel equipment designs for mitigating station blackouts and for enhancing the fuel cycle by achieving more effective waste transmutation. He has been a Chair Professor at Tsinghua University Beijing and a Visiting Professor at the Tokyo Institute of Technology. He received his Ph.D. in Nuclear Engineering from MIT and a master’s degree in Nuclear Physics from the University of Texas Austin. He is also a registered Professional Engineer in Mechanical Engineering in California. His career began with the Reduced Enrichment Research and Test Reactors (RERTR) program. He was the author of COBRA-3C/RERTR. He managed for 23 years Nuclear Analysis Methods at EPRI, comprising reactor physics, thermal hydraulics, severe accidents, and abnormal occurrences and accidents analysis. He was also a manager for issue resolutions to significant safety questions, that include scram reduction, pressurized thermal shock, flow instability, positive moderator temperature coefficient, station blackout, design basis evaluation, steam generator tube rupture, and the Chernobyl accident. Dr. Chao was the project manager for severe accident analysis tools, the MAAP4 and MAAP5 codes, which are now used worldwide. For these codes, he was responsible for the code development, validation, delivery, maintenance, QA, marketing, funding, contract requirements, and user group activities. Other tools he has been responsible for include major thermal hydraulics and reactor physics analysis tools: RETRAN02, RETRAN03, CORETRAN, VIPRE, GOTHIC, COMMIX, ANISN, SIMULATE-E, CASMO, and CPM-3. He has also reviewed PRA projects, material damages caused by radiation, safety analyses of new designs, MHD, and fusion reactor blanket designs. As an editor for the journal Nuclear Engineering and Design of Elsevier for 17 years, he reviewed more than 3,000 manuscripts submitted for publication, covering a complete spectrum of nuclear engineering subjects and topics. Dr. Chao has been the Secretary for the Thermal Hydraulics Division of the American Nuclear Society and a recipient of an American Nuclear Society Presidential Citation. He is the founder of the NUTHOS (Nuclear Thermal Hydraulics, Operations, and Safety) series of international conferences and served the roles as the Technical Co-Chair for NUTHOS-1, General Co-Chair for NUTHOS-4, Technical Co-Chair for NUTHOS-5, Technical Co-Chair for NUTHOS-9, and General Co-Chair for NUTHOS-10.
Plenary Title: Prospect and History of Nuclear Energy
The theme of this presentation is to illustrate “A Complete Perspective of Nuclear Energy.” A complete perspective should include the history and the prospect of all relevant subjects on nuclear energy. Since 1980, many safety issues were surfaced and resolved, including Pressurized Thermal Shock, Anticipated Transient without Scam, and many others. The technical natures of Steam Generator Tube Rupture and Scram Reductions were properly addressed. Yet, Station Blackout remained an alarming scenario and indeed caused a catastrophe at Fukushima in 2011. Probabilistic Risk Assessment became a useful tool for evaluating plant safety. The nationwide effort of Independent Plant Evaluation for all US plants during the 1980s and 1990s turned out to be a fruitful exercise. The TMI-2 incident offered an opportunity for improving safety standards by requiring natural circulation capability for PWRs. The Chernobyl accident led to the ultimate shutdown of all RBMKs due to their intrinsic unsafe property of positive temperature coefficient. The accident also reveals the lack of safety culture and the missing adherence to the operating procedures. The Fukushima accident demonstrates the catastrophe was caused by the lack of safety culture, poor understanding of the severe accident scenarios, and the missing operational guidelines for managing severe accidents. During the last decade, the US repository Yucca Mountain failed to finish the obligated tasks due to political conflicts. The US DOE now adopts the consent-based approach to resolve the gridlock circumstance following the successful strategies on the repository projects in Finland and Sweden. The reprocessing in the US is not moving forward due to several reasons: abundant uranium mines were discovered and the need for recovering plutonium from spent fuels no longer an urgent mission. Studies show that new approaches to fuel cycle strategies could be useful for resolving technical, economic, and political issues related to nuclear wastes and the management of spent fuels. Fast breeder reactors ceased to operate due to technical difficulties and the diminishing need to breed plutonium. Yet, new designs of fast reactors remain an option for their capability of consuming high-level wastes while producing power. Increasing uranium enrichment to 20% through the HALEU endeavor is welcome and expected practice in the near future to furnish the required fuels for use in the new designs of reactors, for better fuel economy and efficient transmutation of nuclear wastes. In recent years, the small modular reactor designs have become a popular choice for their low demand of a heavy upfront investment, better control of manufacturing quality, and construction costs. All vendors have announced their blueprints for their showcase SMR designs to address the flourishing market. The technical basis for the limits on the radiation doses received by a biological entity for healthy considerations has been revisited. The potential health benefits of low-dose radiation have received enough attention to warrant more thorough studies. Decarbonization considerations may be implemented sooner than expected. Instead of taxing the carbon-producing power plants or machines, credits of cash values have been awarded to nuclear power plants in a few U. S. states indicating the carbon taxing mechanisms may be more effective with such practices. Many states changed their position toward nuclear power. A few days ago, California’s prominent stateswoman Dianne Feinstein openly announced that she has changed her mind and is now supportive of nuclear power. Meanwhile, legislation is being discussed in the state assembly for the life extension of the Diablo Canyon nuclear power plant in California. Nuclear power for space travel has gained its momentum by revitalizing the research, development, and design of specific reactor types with practical objectives. In recent months, government agencies have awarded contracts to commercial entities for designing nuclear devices for purposes of space travel. ITER is about to finish with a mission of demonstrating energy break even. The commercial version of the gigantic facility would follow as the next task. NASA announced the success of the Lattice Confinement Fusion, leading to alternative energy sources for space travel and an energy production device on earth. Currently, 60,000 cargo ships are powered by carbon-producing machines and are candidates for converting to nuclear-powered vessels. The concept of a floating nuclear power plant has gained attention in recent years as such portable nuclear power plants could be tugged to the seashores of developed countries for economic development. Such an arrangement could address many related issues, including the upfront investment for power plants, nuclear fuel and waste handling, proliferation concerns, educational qualifications, training for expertise, and local environmental requirements. The landscape for the development of nuclear energy has turned into nothing but a flourishing scene.

Plenary III
Danrong Song
Chief Designer of ACP100 SMR,
Nuclear power Institute of China, China National Nuclear Corporation, China

Graduated and obtained Doctoral degree of Engineering in Nuclear Science and Engineering
Small and medium size reactor overall design. Isotopic production reactor overall design. Seawater desalination, low temperature nuclear heating plant feasibility study. Chief designer of SMR demonstration project ACP100 in China.
Plenary Title: SMR—ACP100 and Its Multiple Application
Multi-purpose modular small pressured reactor ACP100 has been developed by CNNC and started in construction stage last year. Developing nuclear energy is the measure of meeting national economic development and satisfying the need of energy conservation and emission reduction. Large size nuclear power units are hardly suitable to the application of regional network and the non-electric fields.
Small size reactor needs from the great majority of developing countries and the vast Midwest area of China. Domestic cities and regions put forward an urgent request to nuclear energy heating, desalination, and district heating.

Track 1: Fundamental Thermal-Hydraulics
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.

Track 2: Computational Thermal-Hydraulics and CFD Method
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.

Track 4: Safety and Severe Accidents
Dr. Masaki Amaya
Office for Analysis of Regulatory and International Information,
Sector of Nuclear Safety Research and Emergency Preparedness,
Japan Atomic Energy Agency (JAEA), Japan

Dr. Masaki Amaya joined Toshiba Corp. in 1992. He had been on loan to Nippon Nuclear Fuel Development Co. Ltd. from 1992 to 2000 and to Global Nuclear Fuel-Japan (GNF-J) from 2000 to 2002. He moved from Toshiba Corp. to GNF-J in 2002 and had been working until 2003. During these periods, he had been working in R&D fields of nuclear fuel pellet, e.g. pellet fabrication, irradiation behaviors of fuel pellet and fuel rod, etc. He obtained his PhD of engineering in 1998 from Nagoya university in Japan, regarding thermal conductivity change in nuclear fuel pellet.He moved to Japan Atomic Energy Research Institute (currently Japan Atomic Energy Agency (JAEA)) in 2003 and was engaged in nuclear fuel safety research using a research reactor NSRR and a hot laboratory RFEF in JAEA.
He had been dispatched as a secondee to the OECD Halden Reactor Project (HRP) in Norway from 2006 to 2008 to join the international fuel and materials irradiation program using the Halden Reactor. In HRP, he contributed to some irradiation tests of fuel rods and their data analyses.After he returned to JAEA, he has been working in the field of fuel safety. He had acted as the leader of Fuel Safety Research Group in Nuclear Safety Research Center (NSRC) from 2013 to 2020, and also acted as the head of Reactor Safety Research Division in NSRC from 2018 to 2020.His current position is the director of Office for Analysis of Regulatory and International Information in Sector of Nuclear Safety Research and Emergency Preparedness of JAEA. As the director of the Office, he is leading activities for the improvement of the safety of nuclear facilities, e.g. analyses of the Fukushima dai-ichi nuclear power station accident, the application of graded approach to the safety management of nuclear facilities, etc.
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.

Track 5: Thermal-Hydraulics and Safety of Advanced Reactors
Dr. Kyoung-Ho Kang
Korea Atomic Energy Research Institute (KAERI), Korea

Dr. Kyoung-Ho Kang is the Director of Innovative System Safety Research Division in Korea Atomic Energy Research Institute (KAERI). He received his PhD degree in Nuclear and Quantum Engineering from Korea Advanced Institute of Science and Technology (KAIST) in 2009, his master degree in Nuclear Engineering from Seoul National University in 1995. He has been working for Korea Atomic Energy Research Institute (KAERI) since 1995. His research interests are thermal-hydraulic integral effect testing for advanced pressurized water reactor, system-scale safety analysis, and verification of new safety concepts, i.e. passive safety systems. As the Director, he oversees the whole research activity related with the system thermal-hydraulics and safety in KAERI.
Keynote Title: Thermal-Hydraulic Effect Test Validation for Nuclear Safety against Design Extension Conditions by utilizing Passive Safety Systems
As safety concerns regarding severe accidents have been magnified after the Fukushima accident, design extension conditions (DECs) are considered as multiple high-risk failure accidents. Regarding the multiple failure accident, the fuel degradation should be prevented in any case for sustainable nuclear energy. A passive safety system has received widespread attention to reinforce safety against the DECs as well as against design basis accidents (DBAs). KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, ATLAS to resolve safety issues, to verify the performance of new safety systems, and to validate safety analysis codes. Especially, OECD/NEA international joint project has been launched to investigate the safety against DECs by utilizing ATLAS which incorporates various passive safety systems such as passive residual heat removal system and passive core make-up system. A series of tests have been performed in the framework of the OECD/NEA ATLAS project to produce clearer knowledge of the actual phenomena and to investigate the multi-dimensional thermal-hydraulic behaviors. In this study, as for the DECs, a station blackout (SBO) and multiple failure accidents such as SBO with a loss of coolant accident (LOCA) were experimentally investigated. The test results show that a passive safety system has an effective core cooling performance during an accident transient.

Track 6: Plant Operation and Maintenance
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.

Track 8: Small Modular Reactors and Micro Reactors
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.