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Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Kashika Joho Gakkai-Shi, 35(Suppl.2), p.59 - 60, 2015/09
no abstracts in English
Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Nihon Kikai Gakkai Dai-27-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), 2 Pages, 2014/11
no abstracts in English
Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Nihon Kikai Gakkai Dai-26-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.703_1 - 703_2, 2013/11
no abstracts in English
Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Kashika Joho Gakkai-Shi, 32(Suppl.2), p.239 - 240, 2012/10
In the thermal-hydraulic design of supercritical water-cooled reactors, it is required to establish a thermal-hydraulic analysis method which can simulate the heat transfer and fluid flow characteristics of supercritical fluids precisely. A crossing flow is a phenomenon in which the fluid flowing through the inside of a fuel bundle in the perpendicular direction moves between subchannels to a transverse direction. The crossing flow influences the removal heat of a reactor core greatly. In the present study, results of the preliminary crossing flow simulation are described.
Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Kashika Joho Gakkai-Shi, 30(Suppl.2), p.359 - 360, 2010/10
no abstracts in English
Takase, Kazuyuki; Muramatsu, Toshiharu; Seki, Akiyuki; Kitamura, Tatsuaki*; Machida, Hiromu*
Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2009 Koen Rombunshu, p.217 - 218, 2009/11
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Shibata, Mitsuhiko; Kitamura, Tatsuaki*; Kume, Etsuo; Zhe, X.*
Nihon Kikai Gakkai Ryutai Kogaku Bumon Koenkai Koen Rombunshu, P. 227, 2006/10
Fluid flow characteristics in a tight-lattice rod bundle with spacers were analyzed numerically using a newly developed two-phase flow analysis code under the conditions of the full bundle size and non-heated isothermal flow. Conventional analysis methods such as sub-channel codes need composition equations based on the experimental data. In case that there are no experimental data about thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the tight-lattice core. Then the direct numerical simulations with the earth simulator were chosen. The axial velocity and void fraction distributions in a simulated tight-lattice rod bundle changed sharply around a spacer and the interface behavior between water and vapor in a narrow coolant channel was clarified using the predicted quantities. The high prospect was acquired on the possibility of establishment of the thermal design procedure of the advanced nuclear reactors with large-scale direct simulations.
Kume, Etsuo; Kitamura, Tatsuaki*; Takase, Kazuyuki; Ose, Yasuo*
Kashika Joho Gakkai-Shi, 25(Suppl.2), p.369 - 370, 2005/10
no abstracts in English
Takase, Kazuyuki; Kitamura, Tatsuaki*; Kume, Etsuo; Ichimiya, Koichi*; Komada, Ichiro*
Nihon Kikai Gakkai Kanto Shibu Yamanashi Koenkai (2003) Koen Rombunshu, No.030-4, p.77 - 78, 2003/00
no abstracts in English
Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Tamai, Hidesada; Kume, Etsuo; Kitamura, Tatsuaki*
Dai-16-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Yoshishu, 7 Pages, 2002/00
no abstracts in English
Tachimori, Shoichi; *
JAERI-Data/Code 96-030, 116 Pages, 1996/10
JAERI-Data-Code-96-030.pdf:2.45MB
no abstracts in English