Implementation of the heat and mass transfer models for BT and post-BT regions in three-field two-fluid CFD

Abe, Satoshi; Obi, Yoshio*; Satou, Akira; Okagaki, Yuria; Shibamoto, Yasuteru

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03

Liquid film behavior and heat-transfer mechanism near the rewetting front in a single rod air-water system

Wada, Yuki; Le, T. D.; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke

Journal of Nuclear Science and Technology, 57(1), p.100 - 113, 2020/01

https://doi.org/10.1080/00223131.2019.1659872

Ultrasound measurement of upward liquid film flow in vertical pipe

Wada, Yuki; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke; Sagawa, Jun*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.4518 - 4531, 2019/08

Liquid film detection under boiling transition (BT) condition is one of the important issues to develop models on dry out and rewet including physical characteristics of liquid film behavior. Although a heater surface temperature has been often used in previous studies to detect the position of liquid film front, it is difficult to accurately identify the position from the temperature measurement. Therefore, we are developing a nonintrusive measurement technique for detecting thin liquid film thickness under BT and rewet condition using ultrasound. In this study, we focus on high accuracy measurement for liquid film thinner than 0.1 mm by using high frequency ultrasound of 15 MHz and developing a signal processing method. Liquid film measurement results were found to agree with liquid film thickness correlations. Based on a comparison with constant current method, it is concluded that the present technique gives more reasonable liquid film thickness than constant current method.

Experimental investigation of Post-BT heat transfer and rewetting phenomena

Satou, Akira; Wada, Yuki; Le, T. D.; Shibamoto, Yasuteru; Yonomoto, Taisuke

Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 12 Pages, 2018/00

Experiments were performed under the condition of AOO for BWRs to obtain Post-BT heat transfer rate, deposition rates of liquid droplets, and the rewetting behavior after the core dryout. Rewetting behavior was analytically investigated and a relation among the rewetting velocity, the hot wall temperature, and the heat transfer rates in the precursory cooling and wetted regions were obtained. In addition, experiments simulating the condition of ATWS were newly performed with simulated ferrule spacers especially to investigate the spacer effect. It was found that the heat transfer rates were enhanced by the spacers, which were compared with existing prediction models for the validation. The spacers also appeared to increase the rewetting velocity slightly. Since the precursory cooling was found to play an important role on the rewetting behavior through the series of prior experiments, new experiments are conducted focusing on the precursory cooling. In those experiments, the behaviors of liquid film and droplets around the rewetting front were observed to investigate the mechanism of the precursory cooling.

Thermal hydraulic safety research at JAEA after the Fukushima Dai-ichi Nuclear Power Station accident

Yonomoto, Taisuke; Shibamoto, Yasuteru; Takeda, Takeshi; Satou, Akira; Ishigaki, Masahiro; Abe, Satoshi; Okagaki, Yuria; Sun, Haomin; Tochio, Daisuke

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5341 - 5352, 2015/08

RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project

Abe, Satoshi; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo

Journal of Nuclear Science and Technology, 51(10), p.1164 - 1176, 2014/10

https://doi.org/10.1080/00223131.2014.911710

no abstracts in English

A Preliminary 3D steam flow analysis for CET behavior during LSTF SBLOCA experiment using FLUENT code

Irwanto, D.; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo

Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07

https://doi.org/10.1115/ICONE21-16808

Analysis of BWR station blackout accident; Thermal-hydraulic behavior up to severe core damage in Fukushima Daiichi Power Plant Unit 2

Watanabe, Tadashi; Ishigaki, Masahiro; Satou, Akira; Nakamura, Hideo

Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.240 - 244, 2011/12

https://doi.org/10.3327/taesj.J11.022

The analysis of long-term station blackout accident of BWR has been performed using TRAC-BF1 code. The actuation of RCIC was assumed, and the results were compared with the observed data at the Fukushima Daiichi power plant unit 2 reactor. The effectiveness of recovery action for reactor cooling was discussed after the termination of RCIC. A BWR-5 with 1100 MW was analyzed, while the unit 2 was a BWR-4 with 780 MW. The reactor pressure and the core liquid level were, however, in good agreement with the observed data. It was confirmed that the quasi-steady state was continued for a long time by the RCIC actuation. The timing of recovery action, which is composed of depressurization and coolant injection, necessary for the clad temperature being less than 1500 K was studied and compared with the unit 2.

Neutron-coupled thermal hydraulic calculation of BWR under seismic acceleration

Satou, Akira; Watanabe, Tadashi; Maruyama, Yu; Nakamura, Hideo

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 5 Pages, 2010/10

In the BWR subjected to an earthquake, the oscillating acceleration attribute to the seismic wave may cause the variation of the coolant flow rate and void fraction in the core, which might result in the core instability due to the void-reactivity feedback. Therefore, it is important to properly evaluate the effect of the seismic acceleration on the core stability from a viewpoint of plant integrity estimation. In the present study the numerical code analyzing the behavior of nuclear power plant under the seismic acceleration is developed based on the 3-D neutron-coupled thermal hydraulic code, TRAC/SKETCH. The coolant flow is simulated with introducing the acceleration attributed to the earthquake into the motion equation of the flow as external force terms. The analyses are performed on a real BWR4-type nuclear power plant with the sinusoidal acceleration and the acceleration obtained from the response analysis to an actual earthquake. The behaviors of the core and coolant are calculated in the various parameters of acceleration. The effects of the frequency, amplitude and direction of the oscillating acceleration are discussed.

Study on transient void behavior during reactivity initiated accidents under low pressure condition; Development and application of measurement technique for void fraction in bundle geometry

Satou, Akira; Maruyama, Yu; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07

Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 22 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.