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Takeda, Takeshi
JAEA-Data/Code 2024-014, 76 Pages, 2024/12
JAEA-Data-Code-2024-014.pdf:4.0MB
An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.
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.
Kai, Tetsuya; Uchida, Toshitsugu; Kinoshita, Hidetaka; Seki, Masakazu; Oi, Motoki; Wakui, Takashi; Haga, Katsuhiro; Kasugai, Yoshimi; Takada, Hiroshi
Journal of Physics; Conference Series, 1021(1), p.012042_1 - 012042_4, 2018/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
Ito, Kei; Koizumi, Yasuo; Ohshima, Hiroyuki; Kawamura, Takumi*
Mechanical Engineering Journal (Internet), 3(3), p.15-00671_1 - 15-00671_9, 2016/06
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.
Takada, Hiroshi; Naoe, Takashi; Kai, Tetsuya; Kogawa, Hiroyuki; Haga, Katsuhiro
Proceedings of 12th International Topical Meeting on Nuclear Applications of Accelerators (AccApp '15), p.297 - 304, 2016/00
In J-PARC, we have continuously been making efforts to operate a mercury target of a pulsed spallation neutron source with rated power of 1-MW. One of technical progresses is to mitigate cavitation damages at the target vessel front induced by the 3-GeV proton beam injection at 25 Hz. We have improved the performance of a gas micro-bubbles injection into the mercury target, resulting that no significant cavitation damages was observed on the inner surface of target vessel after operation for 2050 MWh with the 300-kW proton beam. Another progress is to suppress the release of gaseous radioactive isotopes, especially tritium, during the target vessel replacement. We have introduced a procedure to evacuate the target system by an off-gas processing apparatus when it is opened during the replacement operation, achieving to suppress the tritium release through the stack. For example, the amount of released tritium was 12.5 GBq, only 5.4% of the estimated amount, after the 2050 MWh operation. After these progresses, the operating beam power for the pulsed spallation neutron source was ramped up to 500-kW in April, 2015.
Ito, Kei; Koizumi, Yasuo*; Ohshima, Hiroyuki; Kawamura, Takumi*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
The authors are developing a high-precision CFD code with an interface tracking method to simulate the gas entrainment (GE) phenomena in sodium-cooled fast reactors (SFRs), which might be caused by a highly-intensified free surface vortex. The GE in SFRs is characterized by an elongated interfacial dent along the vortex core and the bubble pinch-off at the tip of the dent. To simulate this complicated phenomenon, our simulation code has physics-basis algorithms which model accurately the interfacial dynamic behavior, the pressure jump condition at an interface and the surface tension. Several verification problems have been already solved and the accuracy of each individual algorithm is confirmed. In this paper, a basic experiment of the GE is simulated to validate the developed code. The simulation result of the entrained flow rate shows comparable value to the experimental data, that is, our simulation code is considered applicable to the evaluation of the GE in SFRs.
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.
Kojima, Takuji
Shinku, 47(11), p.789 - 795, 2004/11
When flue gas/off gas is irradiated by EB, many free radicals such as OH and active oxygen atom are formed from major components of air: namely nitrogen, oxygen, water and carbon dioxide ecules. The similer reaction can be achieved using UV light and plasma-discharging, but ionizing radiations produce such free radicals at higher density. Such radiation-induced radicals react efficiently with air pollutants, SOx and NOx in coal/oil combustion flue gas at thermal power plants, dioxins in waste incineration flue gas and volatile organic compounds (VOC) even in very low concentration and decompose them into non-toxic substances or change them to removable substances. R & D on EB treatment of flue gas/off gas done in JAERI on the basis of this principle process, as an example, is described in the present paper.
Kojima, Takuji
Shokubai, 46(3), p.248 - 253, 2004/04
The present paper describes research and development on purification technology using electron beams for flue/odd gases containing pollutants: removal of sulfate oxide and nitrogen oxide from flue gases of coal/oil combustion power plants, decomposition of dioxins in waste incineration flue gas, and decomposition/removal of toxic volatile organic compounds from off gas.
Mineo, Hideaki; Goto, Minoru; Iizuka, Masaru*; Fujisaki, Susumu; Hagiya, Hiromichi*; Uchiyama, Gunzo
Separation Science and Technology, 38(9), p.1981 - 2001, 2003/05
Times Cited Count：25 Percentile：65.10(Chemistry, Multidisciplinary)no abstracts in English
Mineo, Hideaki; ; ; Uchiyama, Gunzo; Fujine, Sachio
Proc. of 7th Int. Conf. on Radioactive Waste Mamagement and Environmental Remediation (ICEM '99)(CD-ROM), 3 Pages, 1999/00
no abstracts in English
Mineo, Hideaki; Uchiyama, Gunzo; Hotoku, Shinobu; Asakura, Toshihide; Kihara, Takehiro; ; ; Kimura, Shigeru; ; ; et al.
Proc. of Int. Conf. on Future Nuclear Systems (GLOBAL'99)(CD-ROM), 7 Pages, 1999/00
no abstracts in English
Kihara, Takehiro; Sakurai, Tsutomu*; ; Fujine, Sachio
Proc. of 5th Int. Nucl. Conf. on Recycling, Conditioning and Disposal (RECOD '98), 1, p.830 - 837, 1998/00
no abstracts in English
Nakamura, Hideo; Anoda, Yoshinari; Kukita, Yutaka
Proc. of the Int. Topical Meeting on Safety of Thermal Reactors, p.497 - 503, 1991/00
no abstracts in English
Uchiyama, Gunzo; ; ; ; Sugikawa, Susumu; Maeda, Mitsuru; Tsujino, Takeshi
JAERI-M 90-016, 71 Pages, 1990/02
no abstracts in English
in to dissolver off gas occur?;
UTNL-R-0190, p.17 - 19, 1986/00
no abstracts in English
Sato, Daisuke; Yano, Kimihiko; Kitawaki, Shinichi; Sano, Yuichi; Takeuchi, Masayuki
no journal, ,
