List of Invited Speakers
Plenary 1
November 11 (Mon), 10:10-11:00
Thermal-hydraulics and Severe Accident Research: Challenges and breakthroughs in Nuclear Safety
Ki Yong Choi
Senior Vice-President
Nuclear Safety and Base Technology Lab.
Korea Atomic Energy Research Institute
Korea, Republic of
Short Biography
Dr. Ki Yong Choi is a distinguished expert in thermal-hydraulics and nuclear safety, with significant contributions to the development of next-generation nuclear reactor technologies such as the APR1400 and SMART. He was instrumental in the creation of the SPACE code and served as the Director of Nuclear Thermal-Hydraulic Safety at the Korea Atomic Energy Research Institute (KAERI) from 2012 to 2019. During this time, he helped position KAERI as a global leader in nuclear safety research.
Since 2012, Dr. Choi has also held a faculty position at the University of Science and Technology (UST) - KAERI, where he served as Dean, guiding academic and research programs. He currently serves as the Chair of the Thermal-Hydraulic Division of the Korean Nuclear Society, further strengthening his influence in the field.
Internationally, Dr. Choi is an active contributor to the OECD Nuclear Energy Agency (OECD/NEA), where he serves on both the Working Group on Analysis and Management of Accidents (WGAMA) and the Committee on the Safety of Nuclear Installations (CSNI) bureaus. In these roles, he provides expert reviews and guidance on ongoing and emerging research and development initiatives. Notably, Dr. Choi coordinated the 50th OECD/NEA International Standard Problem (ISP-50) and the OECD-ATLAS project, both of which have been critical in enhancing the safe operation of nuclear power plants among OECD member countries.
Field of Interest or Research
Dr. Ki Yong Choi's research expertise focuses on experimental and analytical studies of thermal-hydraulic phenomena, particularly in the development of advanced models for next-generation light water reactors such as the APR1400, APR+, and SMART. He played a pivotal role in the development of the system-scale safety analysis code SPACE, significantly contributing to the modeling of its heat transfer package. His research interests also extend to the field of research reactor engineering and safety, where he continues to make impactful contributions.
Basic Equations of Two-phase Flow in Nuclear Reactor Safety
Isao Kataoka
Director
Institute of Nuclear Technology, Institute of Nuclear Safety System
Japan
Short Biography
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1973 B. Eng. Kyoto University (Nuclear Engineering)
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1975 M. Eng. Kyoto University (Nuclear Engineering)
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1975 Research Associate, Kyoto University
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1981-1982 Visiting Researcher at Argonne National Laboratory
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1984 D. Eng. Kyoto University (Nuclear Engineering)
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1992 Lecturer, Kyoto University
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1994 Associate Professor, Kyoto University (Dept. Nuclear Eng.)
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1997 Professor, Osaka University, (Dept. Mechanical Eng.)
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2015 Professor, Fukui University of Technology Dean of Faculty of Engineering
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2021 Director, Institute of Nuclear Technology, Institute of Nuclear Safety System
Field of Interest or Research
Heat and fluid flow related to nuclear reactor safety, in particular two-phase flow
Plenary 2
November 11 (Mon), 11:30-12:20
Development of Next-generation Innovative Reactors in Japan
Hidemasa Yamano
Deputy Director
Fast Reactor R&D Department
Japan Atomic Energy Agency
Japan
Short Biography
Dr. Yamano graduated the master course of mechanical engineering from Hiroshima University in 1996, and earned the Ph.D of nuclear engineering from Kyushu University in 2009.
Dr. Yamano joined “Japan Atomic Energy Agency” in 1996 for severe accident study in sodium-cooled fast reactors. Currently, he is deputy director of the fast reactor R&D department for the management of safety design, code & standard, analysis code for licensing, safety-related R&D, in particular severe accident studies and probabilistic risk assessment study of Sodium-cooled Fast Reactor. He is also a coordinator of international collaboration.
Field of Interest or Research
Safety design, safety analysis, severe accident, probabilistic risk assessment, code & standard
V&V of the i-SMR Safety Performance
Han-Gon Kim
President
Innovative SMR Technology Development Agency
Korea, Republic of
Short Biography
Dr Kim graduated from nuclear engineering department of Seoul National University in 1988 and completed his master and Ph.D. degrees in nuclear engineering at KAIST in 1990 and 1993, respectively. Dr Kim worked to develop APR1400 in Korea Hydro and Nuclear Power Co, (KHNP) since 1997. And he had major role to get the standard design approval of APR1400 at 2002. In 2006, he led the national project to develop GEN III+ PWR, APR+.
From 2015, he was the project manager for APR1400 USNRC design certification and EUR certification. From 2020, he was the president of Central Research Institute (CRI) of KHNP and he planned the innovative SMR development with the Korean government.
Field of Interest or Research
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Design of Small Modular Reactor
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Nuclear Thermal Hydraulics
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Nuclear Safety
Keynote 1
November 12 (Tue), 09:30-10:10
Efforts and Challenges toward High-fidelity Experiments and Simulations of Two-phase Flow Heat Transfer in LWRs
Hyungdae Kim
Professor
Nuclear Engineering
Kyung Hee University
Korea, Republic of
Short Biography
Hyungdae Kim is Professor within the Engineering Department at Kyung Hee University (KHU), where he currently leads the human-resource development program entitled in “Next-generation Nuclear Power Plant-based Carbon-neutral Convergence Graduate School” and the director of advanced thermal hydraulics laboratory.
Dr. Kim received the B.S. and Ph.D. degrees in mechanical engineering, Pohang University of Science and Technology, Korea, in 2001 and 2007, respectively. He conducted his postdoctoral research in exploiting and designing nanoscale materials and surfaces for quenching and boiling heat transfer in the department of nuclear science and engineering at MIT, USA, 2008-2010.
Dr. Kim’s research experiences span boiling heat transfer, critical heat flux and quenching in nanofluids, microlayer, wall heat partitioning, single droplet-wall collision heat transfer, fork-end heat pipe for passive cooling of spent fuel pool, and development of multiphase computational fluid dynamics methodology for high void fraction.
Field of Interest or Research
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Boiling heat transfer fundamentals via advanced imaging
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Phosphor thermometry for high-temperature thermal-hydraulics
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Thermal-hydraulic models for high-burnup nuclear fuel cladding
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Hypervapotron technology for ultra-high heat flux cooling
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Severe accident: ex-vessel corium cooling
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Numerical simulation and experiments of helium bubbles in molten salt fast reactor
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Safety regulation for water-cooled SMR
Keynote 2
November 12 (Tue), 09:30-10:10
Effect of Flow Field on Piping Degradation in Nuclear Power Plants
Yoichi Utanohara
Professor
Department of Production Systems Engineering and Sciences
Komatsu University
Japan
Short Biography
He received a PhD degree in in Aeronautics and Astronautics Engineering, University of Tokyo.
He worked for Institute of Nuclear Safety System, Inc. from 2007 and engaged in researches about pipe degradation mechanism and plant safety analysis.
He conducted experimental and numerical studies about flow accelerated corrosion and received a JSME (The Japan Society of Mechanical Engineers) Medal for Outstanding Paper in 2015.
He moved to Komatsu University in 2022.
Field of Interest or Research
His expertise is experimental and numerical studies of thermal fluid engineering related to nuclear power stations, particularly plant aging research and safety analysis.
One of his current research projects is the clarification of the pipe degradation mechanism caused by thermal fatigue.
Keynote 3
November 13 (Wed), 08:30-09:10
Opportunities and Challenges on Multi-physics Experiment, Modeling, and Simulation for Successful Deployment of Molten Salt Reactor
Sung Joong Kim
Professor
Nuclear Engineering
Hanyang University
Korea, Republic of
Short Biography
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B.S. Nuclear Engineering, Hanyang University, Seoul, Republic of Korea (1994-2001)
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M.S. Nuclear Engineering, Seoul National University, Seoul, Republic of Korea (2001-2003)
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S.M. Nuclear Science and Engineering, MIT, Cambridge, USA (2004-2007)
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Ph.D. Nuclear Science and Engineering, MIT, Cambridge, USA (2007-2009)
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Postdoctoral Associate/Research Scientist, MIT NRL, Cambridge, USA (2009-2011)
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Professor, Hanyang University, Seoul, Republic of Korea (2011-Present)
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Deputy Director, Hanyang Institute of Energy and Environment (HY-IEE) (2023-2025)
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Expert Committee Member, Nuclear Safety and Security Commission (NSSC) (2022-2024)
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Part-time Executive, Korea Energy Information and Culture Agency (2024-2026)
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Committee Member: Korean Advanced Reactor Development Program (NEW Clear) (2024-2025)
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Committee Chair, Molten Salt Reactor (MSR) Chapter of NEW Clear (2024-2025)
Field of Interest or Research
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Design Basis Accident analysis of light water reactor using MARS and SPACE codes
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Severe accident analysis including reactor core coolability, hydrogen combustion risk, MCCI, containment failure using MELCOR and CINEMA codes
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Two-phase boiling and condensation heat transfer experiment and modelling
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Critical heat flux focused on surface effect
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Design and performance evaluation of passive safety system of SMR
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Development of natural circulation light water SMR
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Development of on-demand passive flooding safety system applicable for SMR
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Development of Passive Molten salt Fast Reactor (PMFR)
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Multi-physics experiment, modelling, and simulation for MSR
Keynote 4
November 13 (Wed), 08:30-09:10
Development of Advanced Reactors - Activities in MHI and Hitachi-GE
Hideaki Ikeda
Senior Project Manager
Nuclear Energy Systems
Mitsubishi Heavy Industries, Ltd. (MHI)
Japan
Short Biography
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Master of Nuclear Engineering at Osaka University, 1993
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Doctor of Nuclear Engineering at Osaka University, 2002
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Section Manager of Safeguard System Engineering Section in MHI, 2013
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Director of Reactor Core & Safety Engineering Department in MHI, 2020
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Senior Project Manager of Nuclear Energy Systems in MHI, 2024
Field of Interest or Research
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New Reactor Design and Licensing Application
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Improvement of Existing Reactor Design and Operation
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Plant Safety Analysis
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Coupled Neutronics and Thermal-Hydraulics Analysis
Development of Advanced Reactors - Activities in MHI and Hitachi-GE
Takao Kondo
Chief Project Manager
Nuclear Plant Engineering Department
Hitachi-GE Nuclear Energy, Ltd.
Japan
Short Biography
Mr. Takao Kondo joined Hitachi in 1996 and since then has worked in the area of BWR core/fuel design and the relevant licensing analysis. He is also involved in the future reactor development program such as ABWR-II, HP-ABWR and HI-ABWR.
He has an M.S. in Quantum Engineering and System Science from the University of Tokyo.
Field of Interest or Research
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BWR core and fuel, advanced reactor