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Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
JAEA-Data-Code-2023-007.pdf:3.24MB
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
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.
; Kukita, Yutaka
Journal of Nuclear Science and Technology, 33(9), p.696 - 702, 1996/09
Times Cited Count：2 Percentile：24.38(Nuclear Science & Technology)no abstracts in English
; Kukita, Yutaka;
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.131 - 136, 1995/00
no abstracts in English
Maruyama, Yu; Igarashi, Minoru*; Nakamura, Naohiko; Hashimoto, Kazuichiro; Sugimoto, Jun
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering, Vol. 3, 0, p.1247 - 1251, 1995/00
no abstracts in English
Hashimoto, Kazuichiro; Nakamura, Naohiko; Igarashi, Minoru*; Maruyama, Yu; Sugimoto, Jun
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering, Vol. 3, 0, p.1241 - 1246, 1995/00
no abstracts in English
Kumamaru, Hiroshige; Kukita, Yutaka
Int. Conf. on New Trends in Nuclear System Thermohydraulics,Vol. 1, 0, p.119 - 126, 1994/00
no abstracts in English
; Kukita, Yutaka; Yonomoto, Taisuke;
Nuclear Technology, 96, p.202 - 214, 1991/11
Times Cited Count：10 Percentile：71.62(Nuclear Science & Technology)no abstracts in English
Kukita, Yutaka; Nakamura, Hideo; ; ; ; Anoda, Yoshinari; Kumamaru, Hiroshige; Suzuki, Mitsuhiro; ; Yonomoto, Taisuke; et al.
JAERI-M 91-040, 122 Pages, 1991/03
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
The OECD/LOFT Project; Achievements and Significant Results, p.145 - 163, 1991/00
no abstracts in English
Kukita, Yutaka; ; ; ; Anoda, Yoshinari; Kumamaru, Hiroshige; Suzuki, Mitsuhiro; ; Yamamoto, Nobuo; Nakamura, Hideo; et al.
JAERI-M 90-039, 122 Pages, 1990/03
no abstracts in English
Kukita, Yutaka; ; ; Anoda, Yoshinari; Kumamaru, Hiroshige; Suzuki, Mitsuhiro; ; Yamamoto, Nobuo; Nakamura, Hideo; Yonomoto, Taisuke; et al.
JAERI-M 89-225, 117 Pages, 1990/01
no abstracts in English
JAERI-M 8860, 160 Pages, 1980/05
no abstracts in English
JAERI-M 7505, 162 Pages, 1978/02
no abstracts in English
JAERI-M 7239, 177 Pages, 1977/09
no abstracts in English
