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Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2010 Koen Rombunshu, p.344 - 345, 2010/07
no abstracts in English
Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2010 Koen Rombunshu, p.348 - 349, 2010/07
In the thermal hydraulic design of supercritical water-cooled reactor, it is required to establish a thermal hydraulic design method which can precisely predict heat transfer deterioration of supercritical water as the core coolant. Assessments of applicability of turbulence models used in design methods have not been sufficiently performed, since the mechanism of heat transfer deterioration has not been clearly figured out yet. Japan Atomic Energy Agency has started developing prediction method of heat transfer deterioration with large eddy simulation to improve the thermal hydraulic design accuracy. In the present study, simulation results of heat transfer test with Freon are reported.
Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2010 Koen Rombunshu, p.92 - 93, 2010/07
no abstracts in English
Tamai, Hidesada; Nagayoshi, Takuji; Katono, Kenichi; Ito, Takashi; Takase, Kazuyuki
Nihon 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.
Zhang, W.; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2008 Koen Rombunshu, p.108 - 109, 2008/08
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Akimoto, Hajime; Ose, Yasuo*
Nihon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.231 - 232, 2005/08
no abstracts in English
Akimoto, Hajime; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.229 - 230, 2005/08
A thermal-hydraulic research program for Reduced-Moderation Water Reactor (RMWR) has been performed since 2002. The RMWR has a tight-lattice core to attain the breeding of nuclear fuel for the effective use of Plutonium in a light-water reactor system. In this R&D program, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes are being carried out to confirm the cooling performance in tight-lattice fuel assembly of the RMWR. In this paper, outline of the research program is described as well as the latest results of critical power measurement in the large-scale thermal-hydraulic tests and model experiments, which simulates the tight-lattice core of the RMWR.
Yoshikawa, Shinji
Nihon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.311 - 312, 2005/08
None
Onuki, Akira; Shibata, Mitsuhiko; Tamai, Hidesada; Akimoto, Hajime; Yamauchi, Toyoaki*; Mizokami, Shinya*
Nihon Konsoryu Gakkai Nenkai Koenkai 2003 Koen Rombunshu, p.35 - 36, 2003/07
Analytical evaluation of maximum critical power by so-called subchannnel code is indispensable for design of reduced moderation water reactor. In this study, two-phase flow distribution in a tight-lattice rod bundle is investigated using 19-rod bundle experimental rig and subchannnel analysis code NASCA. The flow distribution was measured under so-called churn flow regime and the predictive capability of NASCA was assessed. NASCA can predict the flow distribution qualitatively depending on local pressure drop. Quantitative prediction is also reasonable for liquid phase but the gas phase distribution was underestimated. Void-drift model has a dominant contribution and we should improve the model for the tight-lattice rod bundle.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada; Kano, Takuma; Akimoto, Hajime
Nihon Konsoryu Gakkai Nenkai Koenkai 2003 Koen Rombunshu, p.33 - 34, 2003/00
no abstracts in English
Onuki, Akira; Akamatsu, Mikio*; Akimoto, Hajime
Nihon Konsoryu Gakkai Dai-5-Kai Oganaizudo Konsoryu Foramu Hobunshu, p.87 - 92, 2001/09
Multi-dimensional analyses have been expected with expanding computation resources for gas-liquid two-phase flow in a complex geometry such as fuel rod bundles. Japan Atomic Energy Research Institute is developing a numerical analytical method for the geometry effect, which is based on three-dimensional two-fluid model. In this study, a general curvilinear coordinate system was introduced to the two-fluid model code ACE-3D and air-water two-phase flow around a circular cylinder was analyzed. The present method predicts an air concentration to vortex regions behind the cylinder and a temporal fluctuation of vortex intensity; these two phenomena have been observed in experiments. It is clarified that the phenomena depend on a relative relationship between the drag force and the inertia of bubbles due to pressure fields.
Sobajima, Makoto
Konsoryu Rekucha Shirizu Dai-3-Kai; Kiso Kara Saizensen Made, p.55 - 67, 1989/00
no abstracts in English
Shinozaki, Tatsuya*; Monji, Hideaki*; Kamide, Hideki
no journal, ,
The aim of this study is to reveal the influences of the working fluid properties, the flow configuration and the flow conditions on the characteristics of the free surface vortex. For investigation on the working fluid effect, temperature of the fluid was varied. Flow configuration was altered by changing the water level and by adding structures to the outlet. Then, we investigated the shape of the detached bubble when the surface active agent is added. We found out that all these parameters have considerable influences on the behavior of the free surface vortex.
Oiwa, Hiroshi; Murai, Yuichi*; Yoshikawa, Shinji
no journal, ,
Helically coiled type steam generator is used in Fast Breeder Reactor "Monju". The characteristics of this steam generator is affected by the centrifugal acceleration due to curvature of the heat transfer tube. The centrifugal acceleration forces the liquid phase to the wall of the tube. This phenomenon suppresses the rapid boiling that may induce the hydrodynamic oscillation. In this study, the effect of centrifugal acceleration and inlet conditions of the fluid is investigated by experiment using backlight imaging tomography and numerical simulation using CIP-level set method. We focused on a slug flow regime in which centrifugal acceleration dominates the flow. The present study showed two types of the centrifugal acceleration effects. One is the effect that holds the liquid phase near the tube wall, and another is the liquid-covering effect on the wall by secondary flow.
Nagai, Niro*; Yoshikawa, Shinji
no journal, ,
The helically coiled tube of heat exchanger is used for the evaporator of prototype fast breeder reactor "Monju". This report aims at the grasp of two-phase flow phenomena of forced convective boiling of water inside helically coiled tube, especially focusing on oscillation of dryout point. A shell & tube structure made of glass is used as heat exchanger. Water flows up inside helical tube and the high temperature oil flows down in the outside tube. The oscillation of the dryout point was observed, that is mainly caused by intensive nucleate boiling near the dryout point and evaporation of thin liquid film flowing along the surface. The effects of experimental parameters, such as oil temperature, water temperature, water flow rate, diameter and curvature of the tube, on amplitude and cycle of oscillation were experimentally grasped.
Hattori, Shuji*; Inoue, Fumitaka*; Kurachi, Hiroaki*; Tsukimori, Kazuyuki; Yada, Hiroki
no journal, ,
no abstracts in English
Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Akimoto, Hajime
no journal, ,
Estimation of void fraction in tight-lattice rod bundles was carried out. Five types of void fraction experiments with 7-, 14-, 19- and 37-rod and rod-gap of 1.0-1.3mm bundle and spacer effect tests are being conducted under from the atmospheric pressure to 7.2MPa, and also applicability of the numerical analysis codes and drift-flux model to the tight-lattice rod bundle on void fraction estimation were evaluated. Extensibility of the advanced numerical analysis codes, NASCA, ACE-3D, TPFIT and the system accident analysis code, TRAC-BF1 to the tight-lattice rod bundle was verified and conformed that the codes calculate void fraction distribution qualitatively good agreement with the experimental data.
Ito, Kei; Yamamoto, Yoshinobu*; Kunugi, Tomoaki*
no journal, ,
For the purpose of direct simulation of gas entrainment (GE) phenomena in a sodium-cooled fast reactors, we are developing a high-accuracy seamless physical simulator based on computational scientific methods. In this study, a high-accuracy calculation method for gas-liquid two-phase flow on unstructured mesh has been developed. In this paper, formulations of the calculation method and calculation results of the verification works solving slotted-disk rotation problem are presented. As the verification result, it became clear that the present method had comparable or higher calculation accuracy comparing with conventional high-accuracy methods. In addition, a correction method was introduced to the advection term of the volume fraction transport equation to improve volume preservation characteristics of each phase. This correction method could lead higher calculation accuracy on unstructured mesh comparing with the original method (without correction method).
Ito, Kei; Kunugi, Tomoaki*
no journal, ,
For the purpose of direct simulations for gas entrainment (GE) phenomena, we are developing a high-accuracy gas-liquid two-phase flow simulation method on non-orthogonal meshes. In this study, appropriate formulations satisfying a mechanical balance between surface tension and pressure were derived to achieve a simulation with high-accuracy on distorted meshes. In the formulation, a surface tension potential was introduced. In addition, a new formulation to calculate a pressure gradient was introduced. Finally, the new formulations were verified by calculating a stationary gas-bubble in liquid. As a result, the present method succeeded in eliminating perfectly spurious velocities induced by conventional methods.
Yada, Hiroki; Kurachi, Hiroaki*; Suzuki, Katsuaki*; Tsukimori, Kazuyuki; Hattori, Shuji*
no journal, ,
Research on cavitation erosion in liquid metal is very important to confirm the safety of fast breeder reactor using sodium coolant and to understand cavitation erosion of spallation in the liquid-mercury target of a neutron source. At the erosion test on vibration cavitation in water and in lead-bismuth, erosion rate in lead-bismuth is ten times as fast as in water. In this study, the flowing system apparatus was developed for cavitation inception test using an orifice in liquid metal. The cavitation inception tests were carried out in liquid metal of lead-bismuth and deionized water. We discussed parameters of cavitation inception. Cavitation inception is not dependent on flow velocity both in deionized water and in lead-bismuth. And incipient cavitation number is between 0.8 and 0.6.