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Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
Hosoi, Hideaki*; Yoshida, Hiroyuki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11
Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11
Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.
Ezure, Toshiki; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11
A fundamental water experiment was performed in a cylindrical tank geometry to investigate the influences of fluid viscosity on the cavitation due to sub-surface vortex (vortex cavitation). In order to clarify the influence of fluid viscosity, the fluid temperature was varied from 10 C to 80 C to change the kinetic viscosity () of fluid from 1.310 to 3.710m/s. The occurrences of vortex cavitation were detected by image analysis on digital images of vortex cavitation captured by a digital CMOS camera. Then, the occurrences of vortex cavitation were evaluated from the relation between the yield fraction curves of vortex cavitation and the cavitation factor under several different conditions. The experimental results showed that the influence of was obvious under the large conditions. However, the influence became smaller according to the decrease of .
Toyooka, Junichi; Kamiyama, Kenji; Konishi, Kensuke*; Tobita, Yoshiharu; Sato, Ikken
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
In this study, the heat flux from the molten core materials to the outer surface of the inner duct (the pool-to-duct heat flux) was evaluated based on all the EAGLE-1 in-pile experiments available. Through the evaluation, it was understood that the pool-to-duct heat flux was so high in all the in-pile experiments that the duct wall failed without coolant boiling in its behind. It was also indicated that the presence of steel in the molten core mixture played a key role in this high heat flux. Application of the SIMMER-III code for these tests suggested that some model improvements were necessary to simulate pool-to-duct heat transfer behavior in a mechanistic way.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 9 Pages, 2010/11
In the design of Japan Sodium-cooled Fast Reactor (JSFR), the flow-induced vibration (FIV) arising from the piping geometry and operating condition may occur in the short-elbow pipe. In this study, water experiments using two types of 1/8 scaled elbows with different curvature ratio, r/D = 1.0 and 1.5, were conducted in order to investigate the mechanism of flow fluctuation due to the elbow curvature. The measurements of the velocity fields and pressure fluctuation revealed that the periodic pressure fluctuation of St = 0.56 occurred near the separation region which was formed constantly on the wall of inside.
Yamano, Hidemasa; Tobita, Yoshiharu; Sato, Ikken
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 10 Pages, 2010/11
Morita, Koji*; Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 10 Pages, 2010/11
Kamide, Hideki; Hayafune, Hiroki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
Design study and related researches are carried out for Japan Sodium-cooled Fast Reactor (JSFR). Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are undergoing. Here progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed.
Cheng, S.*; Tanaka, Yohei*; Gondai, Yoji*; Kai, Takayuki*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Fukuda, Kenji*; Yamano, Hidemasa; Suzuki, Toru; et al.
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
Ohno, Shuji; Sugahara, Akihiro*; Oki, Hiroshi*; Ohshima, Hiroyuki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 6 Pages, 2010/11
Parametric analyses were carried out as numerical experiments to clarify the basic characteristics of thermal stratification behavior in upper plenum of fast reactors. The analyses were performed changing hypothetically the analytical conditions of sodium flow rate, sodium temperature, and the size of analyzed upper plenum, by using a commercial CFD code FLUENT Ver. 6.3 and the RNG k- turbulence model. The results provided the suggestion that the averaged sodium ascending velocity in the reactor upper plenum region and the sodium temperature difference before and after the transient initiation would be the dominant factors to determine temperature gradient of thermal stratification interface. Further, it was implied that appropriate spatial mesh arrangement in vertical direction around the stratification interface is significant to obtain the accurate numerical solution of interface temperature gradient.