Rchr
J-GLOBAL ID：201701011938553737   Update date: Jan. 30, 2024

Nakamura Akira

ナカムラ アキラ | Nakamura Akira
Affiliation and department：
Institute of Nuclear Safety System, Inc. Institute of Nuclear Technology

About Institute of Nuclear Safety System, Inc. Institute of Nuclear Technology


Job title： Vice Director
Research field  (1)： Fluid engineering
Research keywords  (1)： Machine Learning
Research theme for competitive and other funds  (1)：
  • 2002 - 2004 Research and Development on Shapes of Non-Vibrational Thermocouple in the Pipe System of Industrial Plants
Papers (43)：
  • Akira NAKAMURA, Takayoshi KUSUNOKI. Development of a Machine Learning Method to Predict the Break Diameter during PWR Loss-of-Coolant Accident. Transactions of the Atomic Energy Society of Japan. 2022. 21. 2. 96-105
  • Koji Miyoshi, Masayuki Kamaya, Yoichi Utanohara, Akira Nakamura. Heat transfer coefficient suitable for thermal fatigue assessment at a T-junction. Nuclear Engineering and Design. 2020. 28. 370. 1-12
  • Koji Miyoshi, Masayuki Kamaya, Yoichi Utanohara, Akira Nakamura. Heat transfer coefficient suitable for thermal fatigue assessment at a T-junction. Nuclear Engineering and Design. 2020. 370. 110916-110916
  • Koji Miyoshi, Akira Nakamura. Study on characteristics of wall temperature fluctuation at a mixing tee with an upstream elbow. American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP. 2018. 4
  • Koji MIYOSHI, Akira NAKAMURA, Yoichi UTANOHARA, Masayuki KAMAYA. Numerical simulation of thermal stress fluctuation at a mixing tee for thermal fatigue problems. Mechanical Engineering Journal. 2018. 5. 4. 18-00272
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MISC (65)：
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Lectures and oral presentations  (9)：
  • DEVELOPMENT OF ESTIMATION METHOD USING MACHINE LEARNING TO EVALUATE BREAK DIAMETER IN PWR LOSS-OF-COOLANT ACCIDENT
    (2023 30th International Conference on Nuclear Engineering (ICONE30) 2023)
  • Numerical simulations of cesium distribution of Fukushima-Daiichi nuclear power plant accident and its uncertainty caused by computational mesh
    (2017 The International Congress on Advances in Nuclear Power Plants (ICAPP2017) 2017)
  • Evaluation of Penetration Length in a Closed Straight Two-inch Branch Pipe Using Numerical Simulations
    (The Ninth Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS9) 2014)
  • SIMULATION OF THERMAL STRIPING AT T-JUNCTION PIPE USING LES WITH SMAGORINSKY CONSTANTS AND TEMPERATURE DIFFUSION SCHEMES
    (The Experimental Validation and Application of CFD and CMFD Codes in Nuclear Reactor Technology, CFD4NRS-4, OECD/NEA and IAEA Workshop 2012)
  • BENCHMARK SIMULATION OF TEMPERATURE FLUCTUATION USING CFD FOR THE EVALUATION OF THE THERMAL LOAD IN A T-JUNCTION PIPE
    (The Seventh Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS7) 2010)
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Work history (4)：
  • 2022/07 - 現在 Institute of Nuclear Safety System, Inc. Institute of Nuclear Technology Vice Director
  • 2017/06 - 2022/06 Institute of Nuclear Safety System, Inc. Institute of Nuclear Technology, Nuclear Power Plant Aging Research Center
  • 1996/08 - 2009/06 Institute of Nuclear Safety System, Inc. Institute of Nuclear Technology, Thermal Hydraulics and Mechanics Group Researcher
  • 2004/04 - 2007/03 Fukui University Graduate School of Engineering Associate Professor
Committee career (24)：
  • 2022/07 - 現在 Japan Welding Engineering Society Nuclear Research Committee
  • 2016/08 - 現在 Atomic Energy Society Japan Standards Committee, System Safety Technical Committee
  • 2017/07 - 2023/01 Japan Welding Engineering Society Planning Committee of Nuclear Research Committee
  • 2020/04 - 2021/03 日本機械学会 北陸信越支部 副支部長
  • 2017/04 - 2019/03 Japan Society of Mechanical Engineers Steering Committee of Power and Energy System Division
Show all
Awards (4)：
  • 2021/06 - Institute of Nuclear Safety System, Inc. Prize of Excellent Research Research of accident estimation for nuclear power plant using machine learning
  • 2017/04 - Japan Society of Maintenology Excellent Paper Award Measurement of Flow Accelerated Corrosion Rate at an Elbow Pipe and Combination Effect of an Upstream Orifice
  • 2015/04 - The Japan Society of Mechanical Engineers JSME Medal for Outstanding Paper Correlation between Flow Accelerated Corrosion and Wall Shear Stress Downstream from an Orifice
  • 2000/04 - Institute of Nuclear Safety System, Inc. Prize of Excellent Research Flow Simulation of Structure in Piping
Association Membership(s) (2)：
ATOMIC ENERGY SOCIETY OF JAPAN ,  THE JAPAN SOCIETY OF MECHANICAL ENGINEERS
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