Prof. Zan Wu
Department of Energy Sciences, Lund University, Sweden
Biography: Zan Wu has been an Associated Professor (universitetslektor) since 2018.02 at the Department of Energy Sciences, Lund University, Sweden. He received his B.S. and Ph.D. from Zhejiang University, Hangzhou, China in 2008 and 2013, respectively. He worked as a postdoctoral fellow, a senior researcher, and then he was promoted as tenured Associate Professor in 2018, all at Lund University. He is currently the head of the heat transfer laboratory. He is leading projects granted by various funding agencies, especially the Swedish Research Council. His research interests are mainly multiphase flow, heat transfer enhancement by micro/nanostructures, boiling and condensation, microreactors, microfluidics and nanofluidics, and renewable energy. These basic research topics can easily find their applications, for example, in heat exchangers (evaporators and condensers) and heat pipes, electronics cooling, MEMS, chemical reactors, biomedical engineering and food engineering. Although Zan’s research works are mainly experimental, he has interests in both experimental and numerical works. He has co-authored more than 130 papers in journals and conferences. He currently serves as associate editor of the Heat Transfer Engineering journal.
Prof. Alexander Vasiliev
Moscow Institute of Physics and Technology, Russia
Title: Application of New Perspective Materials as Advanced Tolerant Fuel Cladding Materials in Nuclear Reactors of Future Generation
Abstract: Currently, the comprehension among the specialists and functionaries throughout the world is getting stronger that the nuclear industry can encounter serious difficulties in development in the case of insufficiently decisive measures to enhance the safety level of nuclear objects and to ensure clean energy and green world. The keen competition with renewable energy sources like wind, solar or geothermal energy takes place presently and is expected to continue in future decades. One of main measures of nuclear safety enhancement could be a drastic renovation of materials used in nuclear industry. The analytical models of high-temperature oxidation of new perspective materials including chromium-nickel-based alloys, zirconium-based cladding with protective chromium coating, FeCrAl alloys and composite claddings on the basis of SiC/SiC in the course of design-basis and beyond-design-basis accidents at nuclear power plants (NPPs) are developed and implemented to severe accident computer running code. The comparison with available experimental data is conducted.
The preliminary calculations of nuclear pressurized water reactor loss-of-coolant accidents with new types of claddings demonstrate encouraging results for hydrogen generation rate and integral hydrogen production. It looks optimistic for considerable upgrade of safety level for future generation NPPs using new fuel and cladding materials. The best ATF cladding candidates and current ATF experiments throughout the world will be discussed in the presentation.
Keywords: Nuclear power plant; Accident-tolerant fuel; Cladding corrosion and oxidation; Hydrogen release; Enhanced level of safety.
Biography: In 1985 Alexander Vasiliev finished Moscow Institute of Physics and Technology (diploma cum laudis). From November 1985 to November 1988 he attended postgraduate courses attached to I.V.Kurchatov Atomic Energy Institute, Moscow. In March 1989 Alexander Vasiliev defended the Ph.D. entitled, "Models of Thermal Processes under the Action of the Moderate Energy Fluxes on Materials". Principal advisor - Prof. L.A.Bolshov, currently Head of Nuclear Safety Institute, Moscow, Academician of Russian Academy of Sciences. From 1988 to 1997 Alexander Vasiliev worked at I.V.Kurchatov Atomic Energy Institute, Moscow. Field of occupation: development and verification of Russian reference “best estimate” computer code SOCRAT on water-water nuclear reactor (VVER) and liquid metal fast reactor (LMFR) accident application; development and verification of Russian new generation sub-channel thermal-hydraulic and neutronic code HYDRA, application of code SOCRAT to normal operation, design-basis-accidents & beyond-design-basis-accidents at nuclear power plants, including fission products release and transport in containment and atmosphere, and spent fuel pool accidents. In 1994 he worked as research associate in Department of Nuclear Engineering, University of Tennessee, Knoxville, US. He performed investigation work on theoretical establishment of critical heat flux and thermal limit values under the conditions of natural low-mass low-pressure two-phase flow following reactor accident. From 1997 to present time Dr. Alexander Vasiliev works at Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE) as senior research associate, head of laboratory, principal research associate. The main areas of interest include analysis of accidents at nuclear power plants (mainly Russian VVER type), radioactive waste operation, application and behavior of new materials in nuclear energy including oxidation of zirconium-based and new perspective accident-tolerant fuel claddings in steam-oxygen-nitrogen gas mixtures.