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Takeda, Takeshi
JAEA-Data/Code 2021-006, 61 Pages, 2021/04
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
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
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.
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Times Cited Count:27 Percentile:91.22(Nuclear Science & Technology)Takeda, Takeshi
JAEA-Data/Code 2014-021, 59 Pages, 2014/11
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.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime
Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.
Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09
When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nagatomi, Hideki; Kaminaga, Masanori; Funayama, Yoshiro
JAERI-Tech 2002-034, 40 Pages, 2002/03
JRR-4, a swimming-pool type research reactor with a thermal power of 3.5MW, attained criticality in July 1998, after replacing its 90% enrichment fuel with a 20% enrichment fuel under the Reduced Enrichment Program. As a part of the program, safety analysis on thermo-hydraulics of the reactor core was conducted on cases including single channel blockage accident. With the conclusion that a certain margin on thermo-hydraulics was necessary, investigation and experiments were carried out with an aim to increase the core flow rate. To increase the core flow, it was carried out to reduce the bypass flow in the core and to increase the primary coolant flow rate from 7m/min to 8m/min. After flow measurements using a mock-up fuel element, flow velocity of the fuel channel was determined as 1.45m/s as opposed to the designed value of 1.44m/s, and the ratio of core flow to total flow was 0.88, exceeding the value 0.86 used for the safety analysis.This report describes the JRR-4 core flow increase plan as well as the results of the channel flow rate measurement
Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo
JAERI-Tech 2001-090, 69 Pages, 2002/01
HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.
Kaminaga, Masanori; Murayama, Yoji; ;
JAERI-Tech 97-043, 63 Pages, 1997/09
no abstracts in English
Inagaki, Yoshiyuki; Kunitomi, Kazuhiko; ; Ioka, Ikuo*; *
Nuclear Technology, 99, p.90 - 103, 1992/07
Times Cited Count:9 Percentile:65.12(Nuclear Science & Technology)no abstracts in English
Iguchi, Tadashi; ; Okabe, Kazuharu*; Sugimoto, Jun; ; Okubo, Tsutomu; Murao, Yoshio
JAERI-M 91-174, 98 Pages, 1991/10
no abstracts in English
Iguchi, Tadashi; Sugimoto, Jun; ; Okubo, Tsutomu; Murao, Yoshio
JAERI-M 91-173, 94 Pages, 1991/10
no abstracts in English
Inagaki, Yoshiyuki; Suzuki, Kunihiro; Ioka, Ikuo*; Kunitomi, Kazuhiko;
Nihon Kikai Gakkai Rombunshu, B, 57(542), p.3520 - 3525, 1991/10
no abstracts in English
Inagaki, Yoshiyuki;
JAERI-M 91-002, 22 Pages, 1991/02
no abstracts in English
Inagaki, Yoshiyuki; Fujimoto, Nozomu; Motoki, Yasuo; Iyoku, Tatsuo; Maruyama, So; Shiozawa, Shusaku
JAERI-M 90-223, 30 Pages, 1990/12
no abstracts in English
*; *; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio
JAERI-M 90-085, 51 Pages, 1990/06
no abstracts in English
Okubo, Tsutomu; Sobajima, Makoto; Iwamura, Takamichi; Onuki, Akira; Abe, Yutaka; Adachi, Hiromichi; Murao, Yoshio
JAERI-M 90-083, 155 Pages, 1990/06
no abstracts in English