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Journal Articles

Experimental investigation of void fraction characteristics for downward steam-water two-phase flow in a large diameter vertical pipe

Katono, Kenichi*; Tamai, Hidesada*; Nagayoshi, Takuji*; Ito, Takashi*; Takase, Kazuyuki

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

no abstracts in English

Journal Articles

Measurement of droplet quality of carryover from free surface using throttling calorimeter

Tamai, Hidesada; Nagayoshi, Takuji; Katono, Kenichi; Ito, Takashi; Takase, Kazuyuki

Proceedings of 7th International Conference on Multiphase Flow 2010 (ICMF 2010) (CD-ROM), 7 Pages, 2010/05

For the design the natural-circulation type BWR that utilizes the FSS concept, the development of a predictive model for the droplets entrained with the steam (carryover) from the free surface is indispensable. In this paper, the droplet quality was measured with a throttling calorimeter that could measure the droplet quality based on the isenthalpic process between wet and superheated steam through the throttle. The measurements were carried out under the conditions of a pressure of 1.5-2.5 MPa. The temperature of the superheated steam after passing through the throttle was confirmed to be strongly related to the quality of the wet steam. A modified model based on the measurements proved to be capable of predicting the droplet quality within the range of the database. Evaluating the droplet quality under BWR conditions validated the feasibility of the design of the natural-circulation type BWR that utilizes the FSS concept.

Journal Articles

Measurement of droplet quality by throttling calorimeter

Tamai, Hidesada; Nagayoshi, Takuji; Katono, Kenichi; Ito, Takashi; Takase, Kazuyuki

Nippon Konsoryu Gakkai Nenkai Koenkai 2009 Koen Rombunshu, P. 2, 2009/08

The characteristics of carryover from free-surface in a natural-circulation BWR are an important subject to be resolved for economic and safe design of the reactor. In this study, droplet quality of the carryover in a test section with 0.12 m in diameter was measured using throttling calorimeter with pressure ranging from 1.5-2.5 MPa. The measured droplet quality increases with decrease in distance from free-surface and with increase in vapor volumetric flux, and these trends are similar to those of previous data.

Journal Articles

Numerical simulation of fluid mixing phenomena in boiling water reactor core using advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Zhang, W.; Takase, Kazuyuki

Nippon Kikai Gakkai Rombunshu, B, 74(742), p.1278 - 1286, 2008/06

no abstracts in English

Journal Articles

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles, 3; Numerical evaluation of fluid mixing phenomena using advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Journal of Power and Energy Systems (Internet), 2(1), p.250 - 258, 2008/00

Journal Articles

Numerical evaluation of fluid mixing phenomena in boiling water reactor using advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 5th Joint ASME/JSME Fluids Engineering Conference (FEDSM 2007) (CD-ROM), 8 Pages, 2007/07

Journal Articles

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles, 4; Numerical evaluation of fluid mixing phenomena using advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

Journal Articles

Numerical simulation of single bubbles rising through subchannels with interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Tamai, Hidesada; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 15 Pages, 2005/10

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 4; Large-scale analysis of water-vapor two-phase flow in rod bundles with TPFIT code using earth simulator

Yoshida, Hiroyuki; Ose, Yasuo*; Kureta, Masatoshi*; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.106 - 114, 2005/06

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 3; Analysis of liquid film falling down on inclined flat plate

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.25 - 31, 2005/03

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 2; Experimental analysis of 2-channel fluid mixing tests

Nagayoshi, Takuji*; Yoshida, Hiroyuki; Onuki, Akira; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.16 - 24, 2005/03

A detailed gas-liquid two-phase flow analysis code based on an advanced interface-tracking method has been developed. It is expected that the developed code would be able to simulate two-phase cross flow behavior within tight-lattice fuel bundles without relying on any empirical correlations. In order to verify the applicability of the code to simulate two-phase cross flow behavior in such situations, numerical analyses of 2-channel model tests were conducted to compare the air slug deformation and separation behavior caused by cross flow through a narrow interconnection between channels. Although the code underestimated the ascending velocity of the slug, the calculated slug deformation and separation behavior were shown to be quite similar to those observed by a high-speed video camera. Moreover the minimum differential pressure between the subchannels through the interconnection, causing channel-to-channel air transfer to occur could be predicted to within 20Pa. However, further studies of modeling and implementation related to the interface-channel wall interaction, such as a contact angle of a gas-liquid interface at the channel wall, are required for prediction improvements. Nevertheless, the qualitative capability of the developed code to simulate two-phase cross flow phenomena was demonstrated.

Journal Articles

Numerical simulation of single bubble behavior in rod bundle with interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Tamai, Hidesada; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.264 - 269, 2004/12

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large scale numerical simulation, 1; Development of a direct analysis procedure on two-phase flow with an advanced interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09

When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.

Oral presentation

Numerical simulation of bubbly flow in square duct using advanced two-fluid model

Yoshida, Hiroyuki; Misawa, Takeharu; Nagayoshi, Takuji*; Akimoto, Hajime

no journal, , 

no abstracts in English

Oral presentation

Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles, 2; Development of 3-dimensional two-phase flow simulation method

Yoshida, Hiroyuki; Tamai, Hidesada; Nagayoshi, Takuji*; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

no journal, , 

no abstracts in English

Oral presentation

Experimental analysis of 2-channel mixing tests under high pressure by using an advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

no journal, , 

no abstracts in English

Oral presentation

Numerical evaluation of fluid mixing in boiling water reactor using advanced interface-tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

no journal, , 

Oral presentation

Oral presentation

Development of design technology on carryover from free-surface in the upper plenum of natural-circulation type BWRs, 2; Measurement of void fraction under free-surface

Nagayoshi, Takuji*; Tamai, Hidesada; Katono, Kenichi*; Nakagawa, Masaki*; Onuki, Akira

no journal, , 

no abstracts in English

Oral presentation

Development of design technology on carryover from free-surface in the upper plenum of natural-circulation type BWRs, 3; Measurement of droplet diameter/velocity distributions

Tamai, Hidesada; Nagayoshi, Takuji*; Katono, Kenichi*; Nakagawa, Masaki*; Onuki, Akira

no journal, , 

no abstracts in English

28 (Records 1-20 displayed on this page)