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Takeda, Takeshi
JAEA-Data/Code 2021-006, 61 Pages, 2021/04
JAEA-Data-Code-2021-006.pdf:2.78MB
An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. 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 SB-PV-09.
Takeda, Takeshi
JAEA-Data/Code 2018-003, 60 Pages, 2018/03
JAEA-Data-Code-2018-003.pdf:3.68MB
Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.
Takeda, Takeshi
JAEA-Data/Code 2015-022, 58 Pages, 2016/01
JAEA-Data-Code-2015-022.pdf:3.31MB
The SB-HL-12 test simulated PWR 1% hot leg SBLOCA under assumptions of total failure of HPI system and non-condensable gas (nitrogen gas) inflow. SG depressurization by fully opening relief valves in both SGs as AM action was initiated immediately after maximum fuel rod surface temperature reached 600 K. After AM action due to first core uncovery by core boil-off, the primary pressure decreased, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before LSC induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after nitrogen gas inflow. Third core uncovery by core boil-off occurred during reflux condensation. The maximum fuel rod surface temperature exceeded 908 K.
Takeda, Takeshi
JAEA-Data/Code 2014-021, 59 Pages, 2014/11
JAEA-Data-Code-2014-021.pdf:5.16MB
Experiment SB-CL-32 was conducted on May 28, 1996 using the LSTF. The experiment SB-CL-32 simulated 1% cold leg small-break LOCA in PWR under assumptions of total failure of HPI system and no inflow of non-condensable gas from ACC tanks. Secondary-side depressurization of both SGs as AM action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after break. Core uncovery started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first LSC. The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery took place before second LSC induced by steam condensation on ACC coolant. The core liquid level recovered rapidly after second LSC. The maximum fuel rod surface temperature was 772 K. The continuous core cooling was confirmed because of coolant injection by LPI system. This report summarizes the test procedures, conditions and major observation.
Takase, Kazuyuki; Akimoto, Hajime
Applied Electromagnetics in Materials, p.177 - 178, 2001/00
no abstracts in English
H.Lee*; H.Seong*; G.Park*; Kumamaru, Hiroshige; Kukita, Yutaka
Proc. of ASME
JSME 4th Int. Conf. on Nuclear Engineering 1996 (ICONE-4), 3, p.41 - 50, 1996/00
no abstracts in English
Onuki, Akira; ; Murao, Yoshio
Journal of Nuclear Science and Technology, 32(3), p.245 - 256, 1995/03
Times Cited Count：1 Percentile：17.53(Nuclear Science & Technology)no abstracts in English
*; Yagi, Junji*; Kumamaru, Hiroshige
JAERI-M 93-199, 48 Pages, 1993/10
no abstracts in English
P.Bazin*; R.Deruaz*; Yonomoto, Taisuke; Kukita, Yutaka
ANS Proc. of the 1992 National Heat Transfer Conf., p.301 - 308, 1992/00
no abstracts in English
Iguchi, Tadashi; *; Iwamura, Takamichi; ; Okubo, Tsutomu; Onuki, Akira; *; Murao, Yoshio
JAERI-M 91-172, 154 Pages, 1991/10
no abstracts in English
Yonomoto, Taisuke; *
Int. J. Multiph. Flow, 17(6), p.745 - 765, 1991/00
Times Cited Count：49 Percentile：87.39(Mechanics)no abstracts in English
; Kukita, Yutaka; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji
Experimental Thermal and Fluid Science, 3, p.588 - 596, 1990/00
Times Cited Count：18 Percentile：76.57(Thermodynamics)no abstracts in English
Kukita, Yutaka; Katayama, Jiro; Nakamura, Hideo; Tasaka, Kanji
Nucl. Eng. Des., 121, p.431 - 440, 1990/00
Times Cited Count：11 Percentile：72.23(Nuclear Science & Technology)no abstracts in English
; Tasaka, Kanji; *; Kukita, Yutaka; Yonomoto, Taisuke
Proc. of the 4th Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Vol. 1, p.206 - 213, 1989/10
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
; Koizumi, Yasuo; Tasaka, Kanji
Nucl.Eng.Des., 92, p.195 - 205, 1986/00
Times Cited Count：1 Percentile：20.71(Nuclear Science & Technology)no abstracts in English
; Tasaka, Kanji; ; ; ; ; ; *;
JAERI-M 85-037, 224 Pages, 1985/03
no abstracts in English
; Tasaka, Kanji; ; ; ;
Nuclear Technology, 70, p.189 - 203, 1985/00
Times Cited Count：2 Percentile：37.51(Nuclear Science & Technology)no abstracts in English
; Murao, Yoshio
JAERI-M 84-131, 223 Pages, 1984/06
no abstracts in English
*; ; Tasaka, Kanji
JAERI-M 83-210, 105 Pages, 1983/11
no abstracts in English
