Effect of seawater on heat transfer without boiling in internally heated annulus

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

https://doi.org/10.3327/taesj.J15.024

no abstracts in English

Measurement of void fraction distribution in air-water two-phase flow in a 44 rod bundle

Liu, 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.

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 1; Outline of the research project

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

The Validation of the detailed two-phase TPFIT code in air-water two-phase flow in an upward vertical square channel

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.

Thermal-hydraulic experiments with sodium chloride aqueous solution

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.

Self-guiding of 100 TW femtosecond laser pulses in centimeter-scale underdense plasma

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

https://doi.org/10.1063/1.2720374

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.

Experimental evaluations of seawater effects on thermal-hydraulic behavior at severe accident, 2; Seawater thermal hydraulic experiments with annular tube test section

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.

Effects of salinity on heat transfer coefficient of forced convective single-phase seawater flow

Nagatake, Taku; Jiao, L.; Uesawa, Shinichiro; Takase, Kazuyuki; Yoshida, Hiroyuki

no journal, , 

Experimental study of seawater effects on thermal-hydraulic behavior for severe accident conditions

Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki

no journal, , 

Effects of seawater salinity on boiling

Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Koizumi, Yasuo; Yoshida, Hiroyuki

no journal, , 

no abstracts in English