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RISS 인기검색어
- 검색키워드
- 저자 : Yuya Matsumoto (검색결과 3 건)
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Yoichi Kinoue,Yuya Matsumoto,Masaki Sakaguchi,Norimasa Shiomi 한국유체기계학회 2021 International journal of fluid machinery and syste Vol.14 No.1
In order to examine the fundamental characteristics of a corner separation in a decelerating cascade flow, an experimental apparatus was made and the three-dimensional separation around a NACA65 blade in a decelerating channel flow was investigated experimentally. Effect of stagger angle of 14 deg. and 10 deg. on the channel flow was focused on. The experimental investigation by the five-hole probe showed that the accumulations of the low energy fluid was seen around the corner part and the overturning flow due to the secondary flow existed for 14 deg. of stagger angle, whereas the accumulations of the low energy fluid were seen at around midspan for 10 deg. of stagger angle. PIV measurement showed that various focus-type separations were seen in the flow for 10 deg. of stagger angle and three-dimensional vortex structure was considered by using a vortex filament.
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Characteristics of Self-Leveling Behavior of Debris Beds in a Series of Experiments
Songbai Cheng,Hidemasa Yamano,Tohru Suzuki,Yoshiharu Tobita,Yuya Nakamura,Bin Zhang,Tatsuya Matsumoto,Koju Morita 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA)and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
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CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS
Cheng, Songbai,Yamano, Hidemasa,Suzuki, TYohru,Tobita, Yoshiharu,Nakamura, Yuya,Zhang, Bin,Matsumoto, Tatsuya,Morita, Koji Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
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