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Uesawa, Shinichiro; Liu, W.; Jiao, L.; Nagatake, Taku; Takase, Kazuyuki; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 15(4), p.183 - 191, 2016/12
no abstracts in English
4 rod bundleLiu, W.; Jiao, L.; Nagatake, Taku; Shibata, Mitsuhiko; Komatsu, Masao*; Takase, Kazuyuki*; Yoshida, Hiroyuki
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10
To contribute the clarification of the Fukushima Daiichi Accident, Japan Atomic Energy Agency (JAEA) has been performed experiments to obtain void fraction distribution data, including detailed bubble information such as bubble velocity and size, in steam-water two-phase flow in rod bundle geometry under high pressure and high temperature condition, focusing on low flow rate at the core natural circulation flow condition after the reactor scram. In this research, experimental apparatus for measuring void fraction distribution in the 44 rod bundle was constructed. To measure the void fraction distribution under high pressure and high temperature condition (up to 2.8 MPa, 232 
C), two wire mesh sensors (WMSs) were installed. To confirm the applicability of the installed WMSs and the measuring system for two-phase flow in rod bundle, experiments in air-water two-phase flow under atmospheric pressure and room temperature were performed. As a result, it was confirmed that the installed WMSs can be applicable to the two-phase flow in rod bundle. Measured results, such as instantaneous and time-averaged void fraction distribution in the rod bundle, average void fraction across the cross section of the flow channel, bubble length and velocity, were also reported.
Jiao, L.; Liu, W.; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Takase, Kazuyuki*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 11 Pages, 2016/10
Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11
Sakka, Taku*; Jiao, L.; Uesawa, Shinichiro; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09
no abstracts in English
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.387 - 388, 2015/06
In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical square channel was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Matos et al. (2004). Comparisons between the experimental and numerical data revealed, in general, good agreement except serious bubble coalescence appeared in numerical simulation.
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Jiao, L.; Liu, W.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa
Proceedings of 15th International Heat Transfer Conference (IHTC 2014) (USB Flash Drive), 11 Pages, 2014/08
In Fukushima Daiichi nuclear disaster, seawater was injected into the nuclear core, which may change the heat transfer characteristics in the reactor pressure vessels (RPV) due to the different physical properties of seawater and pure water. To remove molten fuel from the Fukushima Daiichi Nuclear Power Plants, it is necessary to know the current status of the reactors. Therefore, in this paper, we measured the basic thermal-hydraulic data in an annular tube with a co-axial heater, which includes the heat transfer rate and the pressure drop, using the sodium chloride aqueous solutions and the synthetic seawater as working fluids. The experiments were performed under atmosphere pressure, with the salinity, the fluid mass flux, the inlet temperature and the heat flux used as the parameters. The experimental results and analyses are reported in this paper and the basic influence of the salinity on the heat transfer and the hydraulic characters are proposed.
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2013 Koen Rombunshu, p.107 - 108, 2013/09
In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical circular pipe was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Lucas et al. (2005), who used wire-mesh sensor to obtain a high resolution of the gas fraction data in space as well as in time.
Chen, L.-M.; Kotaki, Hideyuki; Nakajima, Kazuhisa*; Koga, J. K.; Bulanov, S. V.; Tajima, Toshiki; Gu, Y. Q.*; Peng, H. S.*; Wang, X. X.*; Wen, T. S.*; et al.
Physics of Plasmas, 14(4), p.040703_1 - 040703_4, 2007/04
Times Cited Count：37 Percentile：75.36(Physics, Fluids & Plasmas)An experiment for the laser self-guiding studies has been carried out with 100 TW laser pulse interaction with the long underdense plasma. Formation of extremely long plasma channel with its length, about 10 mm, 20 times above the Rayleigh length is observed. The self-focusing channel features such as the laser pulse significant bending and the electron cavity formation are demonstrated experimentally for the first time.
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Koizumi, Yasuo; Yoshida, Hiroyuki
no journal, ,
no abstracts in English
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Takase, Kazuyuki; Yoshida, Hiroyuki
no journal, ,
no abstracts in English
4 bundleLiu, W.; Nagatake, Taku; Jiao, L.; Shibata, Mitsuhiko; Komatsu, Masao*; Takase, Kazuyuki; Yoshida, Hiroyuki
no journal, ,
To improve and validate the prediction accuracy of two - phase codes, Japan Atomic Energy Agency is working on the measurement of void faction distribution in rod bundles with using wire mesh sensors, under high pressure and high temperature conditions (2MPa, 212C). The test section is a 44 rod bundle, in which two three - layer 99 wire mesh sensors are installed at two different axial positions. As the first step of the experiment, to validate the measuring system, we performed experiments in water - air system under atmospheric pressure, with using water and air flow rates as parameters. Void fraction distributions in the sub-channels of the rod bundle were derived in a wide flow pattern from bubbly flow to slug flow. The water flow rate, from the viewpoint of considering the natural circulation after reactor scrum, was lower than 600 kg/m
s. The data will be used to validate the void fraction correlations and two-phase evaluation codes.
Jiao, L.; Takase, Kazuyuki; Liu, W.; Nagatake, Taku; Uesawa, Shinichiro; Yoshida, Hiroyuki; Shibata, Mitsuhiko
no journal, ,
To construct a database for upwards air/water flows in a vertical pipe, extensive measurements of air/water flows in a vertical pipe using the wire-mesh sensor technology were conducted at the thermal fluid dynamic test facility TPTF of the Japan Atomic Energy Agency. The test section is 4m in length and 58mm in inner diameter, two sets of three-layers-WMS were set separately at the 1.15m and 1.65m elevation of the air injection position. Air was injected from the bottom of the pipe through 0.6mm/1mm/2mm diameter nozzles. The obtained data are characterized particularly by their quantity and their detailed information on important two-phase flow parameters (e.g. radial distribution of the void fraction, the gas velocity and the time and cross-section averaged bubble size distribution for different test section heights). In the near future, we would like to use the WMS to measure the detailed two-phase flow in sub-channels of a simulated bundle flow.
Liu, W.; Jiao, L.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa
no journal, ,
In the Fukushima Daiichi Nuclear Power Plant accident, seawater was injected into the reactors to cool down nuclear fuels. Core cooling with seawater has never been assumed and the effect of seawater on heat transfer in core is not clear. Then, effects of seawater on thermal-hydraulic behavior must be investigated to understand the phenomena occurred in the accident and to evaluate current state of the reactor cores. This paper reports thermal-hydraulic experiments and data including heat transfer rate and pressure drop using aqueous sodium chloride solution and manmade seawater in an annular tube, which simulate a whole reactor core. The experiments are performed under atmospheric pressure condition, with salinity, mass flux, inlet temperature and heat flux used as parameters.
Uesawa, Shinichiro; Liu, W.; Jiao, L.; Nagatake, Taku; Takase, Kazuyuki*; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Takase, Kazuyuki
no journal, ,
no abstracts in English
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
no journal, ,
Nagatake, Taku; Jiao, L.; Uesawa, Shinichiro; Takase, Kazuyuki; Yoshida, Hiroyuki
no journal, ,
no abstracts in English
Nagatake, Taku; Jiao, L.; Uesawa, Shinichiro; Takase, Kazuyuki; Yoshida, Hiroyuki
no journal, ,
