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Okagaki, Yuria; Shibamoto, Yasuteru; Wada, Yuki; Abe, Satoshi; Hibiki, Takashi*
Journal of Nuclear Science and Technology, 60(8), p.955 - 968, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Abe, Satoshi; Okagaki, Yuria
Nuclear Engineering and Design, 404, p.112165_1 - 112165_14, 2023/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Dehbi, A.*; Cheng, X.*; Liao, Y.*; Okagaki, Yuria; Pellegrini, M.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03
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：3 Percentile：45.99(Nuclear Science & Technology)Okagaki, Yuria; Yonomoto, Taisuke; Ishigaki, Masahiro; Hirose, Yoshiyasu
Fluids (Internet), 6(2), p.80_1 - 80_17, 2021/02
Abe, Satoshi; Okagaki, Yuria; Ishigaki, Masahiro; Shibamoto, Yasuteru
Proceedings of OECD/NEA Workshop on Virtual CFD4NRS-8; Computational Fluid Dynamics for Nuclear Reactor Safety (Internet), 11 Pages, 2020/11
Okagaki, Yuria; Shibamoto, Yasuteru; Abe, Satoshi
Proceedings of OECD/NEA Workshop on Virtual CFD4NRS-8; Computational Fluid Dynamics for Nuclear Reactor Safety (Internet), 12 Pages, 2020/11
Sun, Haomin; Shibamoto, Yasuteru; Okagaki, Yuria; Yonomoto, Taisuke
Science and Technology of Nuclear Installations, 2019, p.1743982_1 - 1743982_15, 2019/06
Times Cited Count：13 Percentile：82.61(Nuclear Science & Technology)Sun, Haomin; Machida, Shinichi*; Shibamoto, Yasuteru; Okagaki, Yuria; Yonomoto, Taisuke
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
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：28.28(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.
Sun, Haomin; Shibamoto, Yasuteru; Okagaki, Yuria; Yonomoto, Taisuke
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06
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
Kamiji, Yu; Terada, Atsuhiko; Okagaki, Yuria; Hino, Ryutaro
Journal of Nuclear Science and Technology, 51(7-8), p.964 - 967, 2014/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Hydrogen mitigation is one of the important issues for safe storage of spent cesium adsorption vessels at the Fukushima Daiichi NPS because the production and accumulation of hydrogen is induced by the radiolysis of residual water in the vessel. In the present study, an experimental examination was performed using a miniature acrylic vessel to simulate an upper section of the vessel using particle image velocimetry to clarify the internal flow and to verify the analytical results by the FLUENT code. As the result, weak upflow and circulating flow at the stepped section were successfully visualized, and the validity of the analytical results was confirmed by the flow patterns. Additionally, catalytic reaction test was conducted to evaluate its effectiveness. The results showed that the catalyst retains activity under the humid condition assumed in the real vessel.
Yamagishi, Isao; Nagaishi, Ryuji; Terada, Atsuhiko; Kamiji, Yu; Kato, Chiaki; Morita, Keisuke; Nishihara, Kenji; Tsubata, Yasuhiro; Ji, W.*; Fukushima, Hisashi*; et al.
IAEA-CN-211 (Internet), 5 Pages, 2013/01
Since the accident at Fukushima Daiichi Nuclear Power Station, a large amount of radioactive contaminated water has been generated to cool damaged reactor cores. Adsorption of cesium with zeolite-like media was employed for treatment of this contaminated saline water. As spent zeolite media are highly radioactive, their safe storage is a pressing issue. Japan Atomic Energy Agency has extensively conducted R&D on the management of secondary wastes produced by the operation of the treatment system. Subjects on the safe storage of spent zeolites include the analysis of their characteristics and the evaluation of effectiveness of the present safety measures in consideration of decay heat emission and hydrogen generation by water radiolysis as well as durability of vessels exposed to saline. Preliminary results obtained are described in the present paper.
Okagaki, Yuria*; Sugiyama, Hitoshi*; Kato, Naoto*; Hino, Ryutaro
Jidosha Gijutsukai Rombunshu, 43(4), p.949 - 955, 2012/07
Turbulent heat transfer enhancement using periodically arranged ribs processed on the graphite sleeve of a fuel rod aims to increase heat generation density of the fuel rod, which can potentially improve the economics of a block type high-temperature gas-cooled reactor. Rectangular cross section of rib is selected because of easy processing. Furthermore, it is important to find optimum rib topology such as pitch and height to maximize turbulent heat transfer performance. In order to develop a turbulent model as a design tool for rib-roughened fuel rod, numerical analysis has been conducted to verify applicability of an algebraic Reynolds stress model, which is computationally efficient, to non-isotropic turbulent flow in rib roughened channels. The results of the analysis on a square duct with periodically arranged ribs on the bottom wall are shown to reproduce very well the flow characteristics such as flow separation and reattachment behind the ribs. The model is therefore expected to be applicable to rib-enhanced fuel rod design.
Okagaki, Yuria*; Sugiyama, Hitoshi*; Kato, Naoto*; Terada, Atsuhiko; Hino, Ryutaro
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10
Turbulent flow and heat transfer characteristics in a square duct of 100 mm height with 45 degree square ribs of 10 mm height was analyzed numerically by using algebraic Reynolds stress model including the fixed turbulent Prandtl number and algebraic turbulent heat flux models. In this research, analytical results were compared with the experimental and predicted data reported by Bonhoff et al, which were measured and analyzed turbulent flow fields at Reynolds number 50000 based on bulk velocity and duct height by means of a PIV system and a Reynolds stress model. As a result of this study, it was verified that the presented method was able to predict turbulent flow in duct with ribs through the comparison of calculated results with the experimental data.
Okagaki, Yuria*; Sugiyama, Hitoshi*; Hino, Ryutaro; Kato, Naoto*; Terada, Atsuhiko
Nihon Kikai Gakkai Kanto Shibu Burokku Godo Koenkai Koen Rombunshu, p.35 - 36, 2011/09
Turbulent flow and heat transfer characteristics in a square duct of square ribs was analyzed numerically by using algebraic Reynolds stress model. In this research, as the first step of design code development for HTGR heat exchanger equipment, analytical results were compared with the experimental and predicted data reported by Casarsa et al., which were measured by means of a PIV system.
Terada, Atsuhiko; Kamiji, Yu; Hino, Ryutaro; Nishihara, Kenji; Nagaishi, Ryuji; Yamagishi, Isao; Okagaki, Yuria*
no journal, ,
Japan Atomic Energy Agency has extensively conducted R&D on the management of secondary wastes produced by the operation of the system based on the Fukushima Daiichi research and development roadmap. In this paper, to comprehend the temperature distribution and hydrogen diffusion inside the spent zeolite vessel, we report thermal-hydraulic analytical results.
