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Journal Articles

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 6$$times$$6 rod bundle

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12

Journal Articles

Development of numerical simulation method for small particles behavior in two-phase flow by combining interface and Lagrangian particle tracking methods

Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Local gas-liquid two-phase flow characteristics in rod bundle geometry

Xiao, Y.*; Shen, X.*; Miwa, Shuichiro*; Sun, Haomin; Hibiki, Takashi*

Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08

In order to develop constitutive equations of two-fluid model in rod bundle flow channels, experiments of adiabatic air-water upward two-phase flow in 6$$times$$6 rod bundle flow channel were performed. Local flow parameters such as void fraction, interfacial area concentration (IAC) and so on were measured by a double-sensor optical probe. The area-averaged void fraction and IAC data were compared with the predictions from a drift-flux model and an IAC correlation.

Journal Articles

Online coupling of two-phase flow solvent extraction microfluidics with inductively coupled plasma mass spectrometry

Do, V. K.; Yamamoto, Masahiko; Taguchi, Shigeo; Kuno, Takehiko; Surugaya, Naoki

Current Analytical Chemistry, 14(2), p.111 - 119, 2018/00

A direct coupling of two-phase flow solvent extraction microfluidics with ICP-MS for element-selective analysis is successfully established. Two-phase flow in microchannels of two combined glass chips for continuous extraction and back-extraction is stabilized through balancing the pressure by using an external coiled tube that functions as a flow resistor. The difference of fluid flow rate between microchannels and ICP-MS is adjusted by a proposed interface system including T-junction mixer and a switching valve. An online measurement of rhenium is successfully demonstrated. The calibration curve for Re is carried out in the range of 1 $$mu$$g/L to 20 $$mu$$g/L. The limit of detection is 0.2 $$mu$$g/L with a needed sample volume of one milliliter. Total time including extraction, back-extraction, and measurement is less than one hour. The development of the online coupling is a first step towards future applications to the selective measurement of highly radioactive elements.

Journal Articles

Axial flow characteristics of bubbly flow in a vertical large-diameter square duct

Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 14 Pages, 2017/09

An experimental study on the upward bubbly air-water flows in a vertical large-diameter square duct have been performed by using four-sensor probes. The four-sensor probe were applied in the local measurements at 3 axial positions along the flow direction to obtain interfacial area concentration, 3-D bubble velocity vector and bubble diameter. The obtained void fraction, interfacial area concentration, 3-D bubble velocity vector and bubble diameter provided valuable insight into the flow structure and will serve as a valuable database to develop the mechanistic models for interfacial area transport equation sources and sinks.

Journal Articles

Multi-dimensional gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo

Proceedings of 2017 Japan-US Seminar on Two-Phase Flow Dynamics (JUS 2017), 6 Pages, 2017/06

Journal Articles

Measurement of void fraction distribution in steam-water two-phase flow in a 4$$times$$4 bundle at 2 MPa

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki

Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 $$times$$ 4 bundle under 1.6 MPa (202 $$^{circ}$$C), 2.1 MPa (215 $$^{circ}$$C) and 2.6 MPa (226 $$^{circ}$$C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.

Journal Articles

Gas-liquid bubbly flow structure in a vertical large-diameter square duct

Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo

Progress in Nuclear Energy, 89, p.140 - 158, 2016/05

 Times Cited Count:6 Percentile:17.95(Nuclear Science & Technology)

An experimental study was performed on the local structure of upward air-water two-phase flow in a vertical large diameter square duct by using a four-sensor probe. The four-sensor probe method classifying spherical and non-spherical bubbles was applied as a key measurement way to obtain local parameters such as 3-D bubble velocity vector, bubble diameter and interfacial area concentration. Both the local void fraction and interfacial area concentration indicated radial core-peak and wall-peak distributions at low and high liquid flow rates respectively. The 2 components of the bubble velocity vector in the cross-section revealed that there exists a rotating secondary flow in the octant symmetric triangular area and the magnitude of the rotating secondary flow increases with the liquid flow rate. Some of constitutive correlations of drift-flux model and interfacial area concentration are reviewed to study their predictabilities against the present data.

Journal Articles

Numerical simulation of self-priming phenomena in venturi scrubber by two-phase flow simulation code TPFIT

Horiguchi, Naoki; Yoshida, Hiroyuki; Kanagawa, Tetsuya*; Kaneko, Akiko*; Abe, Yutaka*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

From the viewpoint of protecting containment and suppressing diffusion of the radioactive materials at severe accidents of nuclear power plant, it is important to install filtered venting devices to permit release of high pressure pollutant gas to the atmosphere by eliminating radioactive materials in the gas. A Multi Venturi Scrubber System (MVSS) is one of the devices for the filtered venting, and is used to realize filtered venting without any power supply. The MVSS is composed of a "Venturi Scrubbers" part and a "bubble column" part. In the Venturi Scrubbers part of the MVSS, there are hundreds of the Venturi scrubbers (VS). In an operation mode of the MVSS, the radioactive materials are eliminated through the gas-liquid interface from the pollutant gas to the liquid phase of a dispersed flow in the VS and a bubbly flow in the bubble column part. In the VS, the dispersed flow is formed from the liquid, which is suctioned through the hole for suction (called self-priming). In previous studies, an evaluation method to evaluate the liquid flow rate by the self-priming was developed. However, to develop evaluation methods of performance of the VSs, the two-phase flow behavior must be investigated, including droplet size and velocity difference of liquid and gas phases. Two-phase flow behavior in the VS is complicated, and it is difficult to estimate two-phase flow behavior of the VS by only experimental procedures. In this study, to investigate the hydraulic behavior of the VS, we tried to apply a detailed numerical simulation method of two-phase flow to the numerical simulation of the VS. In the simulation, TPFIT developed in JAEA was used as the detailed numerical simulation method. In this paper, we performed the numerical simulation air-water two-phase flow in the of the lab scale VS by the TPFIT, and numerical results were compared with experimental results.

Journal Articles

Development of prediction technology of two-phase flow dynamics under earthquake acceleration, 16; Experimental and numerical study of pressure fluctuation effects on bubble motion

Kato, Yuki; Yoshida, Hiroyuki; Yokoyama, Ryotaro*; Kanagawa, Tetsuya*; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

Journal Articles

Development of evaluation method of liquid flow rate by self-priming phenomena in venturi scrubber

Horiguchi, Naoki; Yoshida, Hiroyuki; Kanagawa, Tetsuya*; Kaneko, Akiko*; Abe, Yutaka*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 5 Pages, 2014/11

Journal Articles

Development of prediction technology of two-phase flow dynamics under earthquake acceleration

Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*

Mechanical Engineering Journal (Internet), 1(4), p.TEP0025_1 - TEP0025_11, 2014/08

Journal Articles

Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. 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 the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

Journal Articles

Large-scale direct simulation of two-phase flow structure around a spacer in a tight-lattice nuclear fuel bundle

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

Computational Fluid Dynamics 2004, p.649 - 654, 2006/00

no abstracts in English

Journal Articles

A Large-scale numerical simulation of bubbly and liquid film flows in narrow fuel channels

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

Proceedings of 2005 ASME International Mechanical Engineering Congress and Exposition (CD-ROM), 8 Pages, 2005/11

no abstracts in English

Journal Articles

Numerical predictions on a large-scale bubbly flow configuration in a minichannel

Takase, Kazuyuki; Yoshida, Hiroyuki; Akimoto, Hajime; Ose, Yasuo*; Aoki, Takayuki*

Nippon Kikai Gakkai 2005-Nendo Nenji Taikai Koen Rombunshu, Vol.7, p.17 - 18, 2005/09

no abstracts in English

JAEA Reports

Numerical analysis of interfacial growth and deformation in horizontal stratified two-phase flow by lattice Boltzmann method

Ebihara, Kenichi

JAERI-Research 2005-004, 121 Pages, 2005/03

JAERI-Research-2005-004.pdf:19.79MB

This report is the JAERI's report version of the doctor thesis by the author. In this report, first, the validity and usefulness of the application of the two-phase fluid model of the lattice-gas method and the lattice Boltzmann method(LBM) are examined. On the basis of the examination, next, the horizontal stratified two-phase flow that is the fundamental and important flow is simulated by the HCZ model which is one of the two-phase fluid model of the LBM. It is seen that the interfacial growth of the HCZ model satisfies the Kelvin-Helmholtz instability theory and reproduces the theoretical two-phase flow regime map of Taitel and Dukler(T-D map). It is found that more superficial flow velocity of the rare phase is necessary in the channel with the narrow width. The HCZ model can also simulate the droplet generation accompanying more complex interfacial phenomena and reproduce the experimental correlation of Ishii and Grolmes in the range of the distribution of the experimental data.

Journal Articles

Numerical analysis of three-dimensional two-phase flow behavior in a fuel assembly

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

WIT Transactions on Engineering Sciences, Vol.50, p.183 - 192, 2005/00

no abstracts in English

Journal Articles

Master plan and current status for feasibility study on thermal-hydraulic performance of reduced-moderation water reactors

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime

Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12

We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.

Journal Articles

Predicted two-phase flow structure in a fuel bundle of an advanced light-water reactor

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

233 (Records 1-20 displayed on this page)