Overview of our laboratory

It is important to understand or evaluate a heat and fluid transport phenomena (thermal-hydraulic phenomena) for performing thermal design of light water reactors (LWRs), formulate accident management plans and design of new safety equipment of LWRs. The present thermal design of nuclear reactors is performed based on the experimental data obtained by mock-up test facilities. However, in such a case, a large amount of fund and human resources are required for construction and maintenance of experimental facilities.
With the progress of the performance of computers and numerical simulation techniques, numerical simulation based on computational fluid dynamics (CFD) is applied to designs in many engineering fields. By applying CFD to designs of the LWRs, reducing or neglecting of the necessities of mock-up test facilities will be achieved. However, application of the CFD to designs of LWRs is very difficult, because the LWR system is very large, and complicated flow, such as gas-liquid two-phase flow and multi-phase flow, appears in the LWRs.
For the establishment of thermal-hydraulics numerical simulation method for large-scale two-phase flow and multi-phase flow, we are developing two-phase flow and multi-phase flow numerical simulation codes for large-scale computers, performing thermal-hydraulics experiments to construct validation databases and developing measurement techniques for performing these experiments. These developed techniques are applied to the research works for improving the safety of the LWRs and considering the decommissioning process of the Fukushima-Daiichi NPS.

News

Call for entries: Recruitment for new graduate in 2020 -Researcher- Detail

T. Misawa, et al., Development of an Analytical Method on Water-Vapor Boiling Two-Phase FlowCharacteristics in BWR Fuel Assemblies Under Earthquake Condition, Nuclear Reactor, Intech, (2012)

K.Takase, et al., Numerical Visualization on Melting and Solidification of Micron-sized Metallic Particles by Laser Irradiation, Quarterly Journal of the Japan Welding Society 、Vol.29, No.3,(2011) pp43-47 (in Japanese)

T. Nakatsuka, et al., Study on Effect of Local Power Distribution of Fuel Assembly on Critical Power of Reduced-Moderation Water Reactor (RMWR), Transactions of the Atomic Energy Society of Japan, Vol 9, No. 2, (2010) (in Japanese)

H. Yoshida, T. Misawa, T. Suzuki and K.Takase, Current Status of Development of Thermal Hydraulic Design Method for Accelerator Driving System in JAEA, Proc. of TCADS1, (2010)

H. Yoshida, H. Hosoi, T. Suzuki and K. Takase, Development of Advanced Two-fluid Model for Boiling Two-phase Flow in Rod Bundles, Proc. of ICONE18, ICONE18-30219, (2010)

T. Nakatsuka, et al., Numerical Simulations on Turbulent Heat Transfer Characteristics of Supercritical Pressure Fluids, IMECE2009-12710, USA, (2009)

K. Takase, et al., Computational Simulations on Melting Process on Fine Metal Powders with Laser Irradiation Welding, IMECE2009-13219, USA, (2009)

T. Misawa, T. Nakatsuka, H. Yoshida, K. Takase, K. Ezato, Y. Seki, M. Dairaku, S. Suzuki, M. Enoeda, Heat Transfer Experiments and Numerical Analysis of Supercritical Pressure Water in Seven-rod Test Bundle, Proc. of NURETH-13, N13P1437 (2009)

Information

Involvement in the Innovative and Viable Nuclear Energy Technology (IVNET) Development Project, "Development of Supercritical Water-Cooled Reactor in GIF Collaboration (Phase-I)", funded by the Ministry of Economy, Trade and Industry (METI) in cooperation with other organizations since FY 2008. Here, our group takes charge of development of a thermal-hydraulic analysis method in SCWR by CFD. Please refer the following WEB site:
http://www.iae.or.jp/KOUBO/innovation/theme/h20/H20_b3.html

Japan science and technology agency, Nnuclear fundamental strategy research initiative project, Strategic nuclear energy joint research program, “Study on the advancement in detailed prediction technology of two-phase flow at addition of earthquake acceleration,” (2010).