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Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12
Times Cited Count：1 Percentile：31.61(Nuclear Science & Technology)Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Extended abstracts of the 2nd Asian Conference on Thermal Sciences (Internet), 2 Pages, 2021/10
For the Japanese sodium cooled fast reactor, a fuel subassembly with an inner duct structure (FAIDUS) was designed to avoid the re-criticality by preventing the large-scale pool formation. In the present study, using the finite volume particle method, the EAGLE ID1 test which was an in-pile test performed to demonstrate the effectiveness of FAIDUS was numerically simulated and the thermal-hydraulic mechanisms underlying the heat transfer process were analyzed.
Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01
Times Cited Count：5 Percentile：65.59(Nuclear Science & Technology)Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04
Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m
on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.
Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
Maruyama, Shinichiro*; Watatani, Satoshi*
Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10
It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.
Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Abe, Yuta; Sato, Ikken; Ishimi, Akihiro; Nakagiri, Toshio; Nagae, Yuji
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm 
107 mm 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
JAERI-Research 2005-014, 170 Pages, 2005/06
JAERI-Research-2005-014.pdf:7.64MB
A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.
Mishima, Kaichiro*; Saito, Yasushi*
JAERI-Tech 2002-014, 83 Pages, 2002/03
JAERI-Tech-2002-014.pdf:6.83MB
no abstracts in English
Iguchi, Tadashi; Iwaki, Chikako*; Anoda, Yoshinari
JAERI-Research 2001-060, 91 Pages, 2002/02
JAERI-Research-2001-060.pdf:6.34MB
no abstracts in English
Maruyama, Yu; Sugimoto, Jun
Journal of Nuclear Science and Technology, 36(10), p.914 - 922, 1999/10
Times Cited Count：3 Percentile：28.69(Nuclear Science & Technology)no abstracts in English
Onuki, Akira; Okubo, Tsutomu; Akimoto, Hajime
Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00
no abstracts in English
Maruyama, Yu; Sugimoto, Jun; Yamano, N.; Hidaka, Akihide; Kudo, Tamotsu; Soda, Kunihisa
Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants, p.23.5-1 - 23.5-6, 1992/00
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
Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio; *
Thermal Hydraulics of Advanced Nuclear Reactors, p.31 - 38, 1990/00
no abstracts in English
; ; ; ; ; Murao, Yoshio
Proc.2nd Int.Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, p.2 - 104, 1986/00
no abstracts in English
Sudo, Yukio; *; ; ; Kaminaga, Masanori
JAERI-M 85-126, 95 Pages, 1985/09
no abstracts in English
) in the Case of a Blower Trip; ;
JAERI-M 84-174, 43 Pages, 1984/09
no abstracts in English
