Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru
Nuclear Engineering and Design, 437, p.114020_1 - 114020_14, 2025/06
Times Cited Count：0Post-boiling transition (post-BT) heat transfer is essential for analyzing the duration of surface dryout and peak cladding temperature during abnormal transients and accidents in light water reactors. The rewetting phenomenon is very important for evaluating the dryout duration. However, due to the lack of an experimental database on rewetting velocities under high flow and heat flux conditions, sufficient data for model development and validation do not exist. Therefore, a database on rewetting velocities caused by stepwise boundary condition changes under a wide range and multiple combination of thermal-hydraulic conditions was obtained using a single-tube experimental apparatus. Based on this database and the characteristics of rewetting velocities obtained, an experimental correlation for rewetting velocity was proposed. This correlation predicts the rewetting velocity accurately by taking the change in the mass flux of the liquid or gas phase with stepwise transients as a parameter. This suggested that the change in the mass flux of the gas or liquid phase near the liquid film front has a strong influence on the rewetting under extremely high mass flux conditions compared to the reflooding process.
Satou, Akira; Hibiki, Takashi*; Ikeda, Ryo; Shibamoto, Yasuteru
Progress in Nuclear Energy, 180, p.105593_1 - 105593_11, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)During a loss-of-coolant accident in a pressurized water reactor (PWR), there is a risk that pressurized thermal shock (PTS) may occur on the internal wall of the reactor pressure vessel (RPV) due to the flow of emergency core cooling (ECC) water injected into the cold leg that flows into the downcomer. PTS is caused by the rapid cooling of the downcomer wall by the ECC water and is strongly influenced by the temperature of the ECC water, the collision position and velocity of the water jet on the wall, the velocity of the liquid film on the wall, the thickness of the liquid film, and the spread of the downward flow. Therefore, the flow of ECC water discharging from the cold leg to the downcomer may strongly impact PTS events. To help understand this flow phenomenon, we reviewed studies on free outflow from a circular pipe. Experimental findings on the classification of flow conditions, transition conditions between flow conditions, end depth ratio, free surface profile of flow in the circular pipe, and shape of the nappe flowing out from the pipe have been obtained in a form that is almost consistent with each other. In contrast, when considering the flow from the cold leg to the downcomer, it is necessary to deal with the flow field in a specific situation, such as the flow into a narrow gap rather than a free space, the existence of rounded corners at the outlet of the circular pipe, and the influence of steam flow flowing from the core to the cold leg. However, few previous studies consider these factors, so we summarized them as knowledge that needs to be accumulated in the future.
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
Abe, Satoshi; Okagaki, Yuria; Satou, Akira; Shibamoto, Yasuteru
Annals of Nuclear Energy, 159, p.108321_1 - 108321_12, 2021/09
Times Cited Count：5 Percentile：50.16(Nuclear Science & Technology)Satou, Akira; Sagawa, Jun*; Sun, Haomin; Shibamoto, Yasuteru; Yonomoto, Taisuke
Nuclear Engineering and Design, 379, p.111234_1 - 111234_7, 2021/08
Times Cited Count：1 Percentile：10.40(Nuclear Science & Technology)Multi-sensor void probe is efficient to measure the local parameters of two-phase flow such as bubble interface velocity. In general, meniscus due to surface tension is formed and shape of the gas-liquid interface deforms when an object contacts the gas-liquid interface. The deformation of the interface by penetration of the front sensor affects the penetration time of the rear sensor, and as a result, an error can occur in the measurement of the bubble interface velocity. The characteristics of the meniscus formation around the sensor was investigated and the error in the measurement of the interface velocity was evaluated. It was clarified that the size, shape of the sensor and the contact angle of the sensor surface affect the error in interface velocity measurement as well as the interface velocity itself, and no measurement error would occur in air-to-water penetration by using a sensor with a large surface contact angle. A 4-sensor void probe was applied to bubbly flow to measure the bubble interface velocity. The measurement error due to the meniscus also occurred in the actual experimental measurement. It was shown that it is necessary to use only the velocity of the lower surface of the bubble or to make certain appropriate error correction for the velocity of the upper interface of the bubble.
Wada, Yuki; Le, T. D.; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke
Journal of Nuclear Science and Technology, 57(1), p.100 - 113, 2020/01
Times Cited Count：8 Percentile：54.22(Nuclear Science & Technology)Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru; Yonomoto, Taisuke
Nuclear Engineering and Design, 354, p.110164_1 - 110164_10, 2019/12
Times Cited Count：13 Percentile：73.65(Nuclear Science & Technology)JAEA has conducted a series of experimental researches on the Post-boiling transition heat transfer, transient critical heat flux and rewetting for BWRs. Experimental data bases covering the anticipated operational conditions was developed; the significance of the precursor cooling was identified. This paper presents approaches of the present research focusing on the anticipated transient without scram, effects of the spacer and physical understanding of the phenomena for development of mechanistic model together with promising results obtained so far.
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.
Wada, Yuki; Le, T. D.; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 10 Pages, 2018/07
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.
Yonomoto, Taisuke; Shibamoto, Yasuteru; Satou, Akira; Okagaki, Yuria
Journal of Nuclear Science and Technology, 53(9), p.1342 - 1352, 2016/09
Times Cited Count：3 Percentile：25.94(Nuclear Science & Technology)Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences (AOOs) for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. The present study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was firstly defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis.
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
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
Times Cited Count：15 Percentile：71.78(Nuclear Science & Technology)After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Abe, Satoshi; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo
Journal of Nuclear Science and Technology, 51(10), p.1164 - 1176, 2014/10
Times Cited Count：5 Percentile：34.81(Nuclear Science & Technology)no abstracts in English
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Irwanto, D.; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
Nakamura, Hideo; Takeda, Takeshi; Satou, Akira; Ishigaki, Masahiro; Abe, Satoshi; Irwanto, D.
Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 21 Pages, 2013/05
Watanabe, Tadashi; Ishigaki, Masahiro; Satou, Akira; Nakamura, Hideo
Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.240 - 244, 2011/12
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.
Satou, Akira; Watanabe, Tadashi; Maruyama, Yu; Nakamura, Hideo
Progress in Nuclear Science and Technology (Internet), 2, p.120 - 124, 2011/10
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. 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. The coolant flow in the core is simulated with introducing the oscillating acceleration attributed to the earthquake motion into the motion equation force terms. The analyses are performed on a real BWR4-type nuclear power plant with the sinusoidal acceleration and the acceleration obtained from a real 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.
Satou, Akira; Maruyama, Yu; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 5(3), p.263 - 278, 2011/04
A new model for the occurrence of the net vapor generation was developed to improve the predictive capability of best-estimate thermal hydraulic codes for transient void behavior under fast transient condition such as reactivity initiated accidents (RIA). It was clarified that the concept of vapor condensation in the model needed to be improved by analyzing the RIA simulation experiments, thus, the new model for the net vapor generation was developed by using the thickness of thermal boundary layer as a characteristic length of vapor condensation. The new model was introduced into TRAC-BF1 code and was applied to the analyses for the high pressure experiments, confirming that the predictive capability of the modified code was improved.
