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Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00
Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio
Journal of Nuclear Science and Technology, 35(12), p.943 - 951, 1998/12
Times Cited Count：24 Percentile：85.13(Nuclear Science & Technology)no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1815 - 1822, 1997/00
In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and nonuniform heat flux condition. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed.
Matsubayashi, Masahito; Sudo, Yukio; Haga, Katsuhiro
Proc. of ASME
JSME 4th Int. Conf. on Nuclear Engineering 1996 (ICONE-4), 1(PART B), p.699 - 705, 1996/00
no abstracts in English
Onuki, Akira
JAERI-M 92-150, 134 Pages, 1992/10
no abstracts in English
Onuki, Akira; ; Murao, Yoshio
Journal of Nuclear Science and Technology, 29(3), p.223 - 232, 1992/03
no abstracts in English
Sudo, Yukio; *; Kaminaga, Masanori
JSME Int. J., Ser. 2, 34(2), p.169 - 174, 1991/00
no abstracts in English
*; Kaminaga, Masanori; Sudo, Yukio
JAERI-M 88-134, 26 Pages, 1988/07
no abstracts in English
; ; Murao, Yoshio
Nucl.Eng.Des., 107, p.283 - 294, 1988/00
Times Cited Count：65 Percentile：97.64(Nuclear Science & Technology)no abstracts in English
JAERI-M 85-219, 19 Pages, 1986/01
no abstracts in English
Abe, Yutaka; ; Murao, Yoshio
Journal of Nuclear Science and Technology, 23(5), p.415 - 432, 1986/00
Times Cited Count：4 Percentile：48.02(Nuclear Science & Technology)no abstracts in English
Journal of Nuclear Science and Technology, 22(9), p.723 - 732, 1985/00
Times Cited Count：5 Percentile：59.78(Nuclear Science & Technology)no abstracts in English
Sudo, Yukio; *; ; Kaminaga, Masanori;
Journal of Nuclear Science and Technology, 22(8), p.604 - 618, 1985/00
Times Cited Count：59 Percentile：97.81(Nuclear Science & Technology)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
*; Tasaka, Kanji
JAERI-M 9476, 60 Pages, 1981/05
no abstracts in English
; ;
JAERI-M 9080, 77 Pages, 1980/09
no abstracts in English
; ; ; ; ;
Nihon Genshiryoku Gakkai-Shi, 19(6), p.408 - 419, 1977/06
Times Cited Count：0no 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.
