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Onuki, Akira; Akimoto, Hajime
Journal of Nuclear Science and Technology, 38(12), p.1074 - 1080, 2001/12
Multi-dimensional analyses have been expected recently with expanding computation resources for gas-liquid two-phase flow analyses of advanced nuclear systems such as passive safety systems and natural-circulation-type reactors. However, the applicability of previous constitutive equations for multi-dimensional analyses has not been fully investigated especially for the effects of flow path scale because the equations have been assessed for small-scale experiments. In this study, we analyzed the scale effects by the multi-dimensional two-fluid model code using data in 38 mm and 200 mm diameter pipes. We clarified a key-parameter to model the scale effects and developed models for the effects on phase distribution. The scale effects can be classified by the relative relationship between bubble diameter db and turbulent length scale lT. Bubble-induced turbulence is increased under that db is smaller than lT and bubble coalescence is predominated rather than breakup under that lT is about three times larger than db and under higher void fraction. Based on these findings, we established new models for bubble turbulent diffusion and bubble diameter. The applicability was promising through assessments against the 38 mm and 200 mm pipes under different flow rates and against a database for developing flow along 480 mm pipe.
Onuki, Akira; Akimoto, Hajime
Proceedings of the 8th International Symposium on Flow Modeling and Turbulence Measurements (FMTM2001) (CD-ROM), 7 Pages, 2001/12
Multi-dimensional analyses have been expected with expanding computation resources for gas-liquid two-phase flow. We recently developed models for bubble turbulent diffusion and bubble diameter to predict the phase distribution by a multi-dimensional two-fluid model. This study was performed to verify our model. The verification was performed using databases under diameter; 9 mm to 155 mm, pressure; atmospheric to 4.9 MPa, flow rate; superficial gas velocity = 0.01 to 5.5 m/s and superficial liquid one = 0.0 to 4.3 m/s, fluid combination; air-water or steam-water. Through the assessments, our model was found to be applicable to the wide range of flow conditions including the effect of pipe diameter. The shape of phase distribution and the average void fraction are predicted well qualitatively and quantitatively. Since the model is established using the ratio of bubble diameter to eddy size as a key-parameter, the ratio is one of important parameters to develop the constitutive equations in the multi-dimensional two-fluid model.
Onuki, Akira; Kamo, Hideki*; Akimoto, Hajime
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1670 - 1676, 1997/00
no abstracts in English
Nakamura, Hideo
JAERI-Research 96-022, 135 Pages, 1996/05
JAERI-Research-96-022.pdf:3.86MB
no abstracts in English
Nakamura, Hideo; Kukita, Yutaka; *
Journal of Nuclear Science and Technology, 32(7), p.641 - 652, 1995/07
Times Cited Count：5 Percentile：49.5(Nuclear Science & Technology)no abstracts in English
; Anoda, Yoshinari; Kukita, Yutaka
Proc. of 2nd Int. Conf. on Multiphase Flow (ICMF)95-KYOTO,Vol. 2, 0, p.P1_97 - P1_102, 1995/00
no abstracts in English
Onuki, Akira; ; Sudo, Yukio
Proc. of the 2nd Int. Conf. on Multiphase Flow 95-Kyoto, 0, p.FT1.17 - FT1.23, 1995/00
no abstracts in English
Nakamura, Hideo; Kukita, Yutaka; *
Journal of Nuclear Science and Technology, 31(2), p.113 - 121, 1994/02
Times Cited Count：1 Percentile：17.88(Nuclear Science & Technology)no abstracts in English
; Anoda, Yoshinari; Kukita, Yutaka
Nihon kikai Gakkai Dai-72-Ki Zenkoku Taikai Koen Rombunshu, Vol.II, 0, 3 Pages, 1994/00
no abstracts in English
Katayama, Jiro; Nakamura, Hideo; Kukita, Yutaka
Proc. of the Int. Conf. on Multiphase Flows 91-TSUKUBA,Vol. 1, p.7 - 10, 1991/00
no abstracts in English
Anoda, Yoshinari; Kukita, Yutaka; Nakamura, Hideo; Tasaka, Kanji
Proc. on 1989 National Heat Transfer Conf., Vol. 4, 8 Pages, 1989/00
no abstracts in English
; ; Murao, Yoshio
Nucl.Eng.Des., 107, p.283 - 294, 1988/00
Times Cited Count：64 Percentile：97.59(Nuclear Science & Technology)no abstracts in English
Osakabe, Masahiro; Koizumi, Yasuo; ; ; Tasaka, Kanji
Nucl.Eng.Des., 98, p.69 - 76, 1986/00
Times Cited Count：3 Percentile：40.89(Nuclear Science & Technology)no abstracts in English
; Murao, Yoshio
JAERI-M 84-131, 223 Pages, 1984/06
no abstracts in English
*;
Journal of Nuclear Science and Technology, 19(12), p.985 - 996, 1982/00
Times Cited Count：43 Percentile：95.52(Nuclear Science & Technology)no abstracts in English
Koizumi, Yasuo; Yamaji, Tatsuya*; Yamazaki, Kohei*; Otake, Hiroyasu*; Hasegawa, Koji*; Onuki, Akira*; Kanamori, Daisuke*
no journal, ,
Experiments of condensing counter-current two-phase flow in a vertical pipe were performed. This study was intended to examine water accumulation in the up-flow side of steam generator U-tubes of a PWR during the reflux cooling stage of a small break LOCA. It has been apprehended that the water accumulation may result in temporary core liquid level depression. The inner diameter and the length of a test flow channel used in the experiments were 18 mm and 4 m, respectively. The experiments were performed by using steam and water at 0.1 MPa. Two kinds of experiments were conducted; visualization experiments by using a transparent test section and quantitative water accumulation evaluation experiments by using a brass test section. Even if water on the inner surface of the test pipe could not flow downward at the lower portion of the test pipe, a part of water became to flow downward at the upper portion of the test pipe since steam velocity decreased because of condensation. Thus, two-phase mixture level was formed in the upper portion of the test pipe, which resulted in the water accumulation in the pipe. The model to predict the water accumulation was proposed. It predicted the water accumulation reasonably well.
