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Tani, Akihiro*; Shimakawa, Yoshio*; Kubo, Shigenobu*; Fujimura, Ken; Yamano, Hidemasa
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Futagami, Satoshi; Hayafune, Hiroki; Fujimura, Ken; Sato, Mitsuru*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9169_1 - 9169_8, 2009/05
The major target performance of the SG for commercialized FBR is not only economic performance but also property protection performance. The straight double wall tube SG is one of the SG candidates for commercialized FBR, and other SG concepts were studied in this paper. In proposing an alternative SG, alternative technological measures with a double wall tube were investigated and included reinforcing the tube against wastage and quick detection of initial tube leaks. Alternative SG concept candidates for preventing tube failure propagation and mitigation of water leak accidents were proposed through a combination of technological measures. At the end of JFY 2010, the straight double wall tube SG will be decided upon as the result of R&D activities, and alternative SGs evaluated in feasibility studies. A plan for studying feasibility with the technological issues of the alternative SG was proposed.
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Journal of Power and Energy Systems (Internet), 2(2), p.456 - 466, 2008/00
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.85 - 88, 2007/06
As a candidate for next generation reactor, the innovative FLexible-fuel-cycle Water Reactor (FLWR) adopts a remarkably tight triangular lattice arrangement with about 1 mm gap spacing between adjacent fuel rods. In relation to its design, this study presents a statistical evaluation of numerical simulation results of a detailed two-phase flow simulation code (named TPFIT). In order to make clear mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is used to simulate cross flow between two modeled subchannels. Attention was focused on instantaneous fluctuation characteristics of differential pressure between two subchannels and gas/liquid mixing coefficients. With the calculation of correlation coefficients between the differential pressure and gas/liquid mixing coefficients, the time scales of cross flow, e.g. lag times were evaluated, and the effects of mixing section length, flow pattern and gap spacing on correlation coefficients were extensively investigated. The difference in mechanism between gas and liquid cross flows was pointed out.
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
The innovative Water Reactor for FLexible fuel cycle (FLWR) adopts a tight triangular lattice arrangement with about 1 mm gap between adjacent fuel rods. In view of the importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study numerically simulated steam-water two-phase cross flow between two modeled subchannels of tight-lattice rod bundle for the FLWR by using a detailed two-phase flow simulation code with an advanced interface tracking method (named TPFIT), statistically evaluated the simulation results, and clarified mechanisms of cross flow for developing a model. The effects of flow pattern, inlet and outlet of mixing section, and gap spacing on cross flow, and the local and general characters of cross flow were extensively investigated.
Chitose, Hiromasa*; Hotta, Akitoshi*; Onuki, Akira; Fujimura, Ken*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
no abstracts in English
Kiuchi, Kiyoshi; Ioka, Ikuo; Tanabe, Makoto*; Nanjo, Yoshiyasu*; Ogawa, Hiroaki; Ishijima, Yasuhiro; Tsukatani, Ichiro; Ochiai, Takamasa; Kizaki, Minoru; Kato, Yoshiaki; et al.
JAEA-Research 2006-023, 173 Pages, 2006/03
JAEA-Research-2006-023.pdf:20.51MB
The research concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100 GWd/t of BWR was pursued for 5 years from 2001 to 2005. On the Phase 1, the modified stainless steel of Fe-25Cr-35Ni-0.2Ti as fuel claddings and Nb-Mo alloy as a liner for inhibiting the pellet- clad interaction were selected as candidate materials, by evaluating fundamental properties required to BWR cladding materials, that are the nuclear economy, radioactivity, mass-transfer, irradiation properties, mechanical properties so on. On the present study, the making process of cladding tubes, lining by diffusion bonding, end plug by laser welding were developed and optimized, by considering the practical use of fuel elements consists of these candidates. The practical applicability was basically examined by irradiation tests using the accelerator of TIARA and the research reactor of JRR-3, for mainly confirming the resistance to IGSCC as one of the current important issues of BWR core materials of low carbon grade stainless steels. Creep and fatigue testing data were also obtained for evaluating the long performance of candidate materials. The behavior as fuel elements was analyzed with the safety calculation code for BWRs. The obtained results were established as a data base system, by considering the applicability to the fuel design and in-pile loop tests.
Onuki, Akira; Tamai, Hidesada; Yoshida, Hiroyuki; Shibata, Mitsuhiko; Akimoto, Hajime; Chitose, Hiromasa*; Hotta, Akitoshi*; Fujimura, Ken*
no journal, ,
no abstracts in English
Ioka, Ikuo; Ishijima, Yasuhiro; Kiuchi, Kiyoshi; Kizaki, Minoru; Kato, Yoshiaki; Fujimura, Ken*; Obata, Hiroyuki*
no journal, ,
The IASCC and loss of ductility concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100GWd/t of BWR was performed after neutron irradiation. The specimens machined from tubes made of Fe-25Cr-35Ni-0.2Ti UHP and SUS304 were irradiated at 290
C up to 1.8dpa(1.5
10
n/m
) in JRR-3. The ultimate tensile strength of both specimens increased and the fracture elongation decreased in tensile test at 288C. The loss of ductility was almost equal with data of previous literatures. IASCC was recognized on SUS304 by SSRT testing under high temperature water, but not on Fe-25Cr-35Ni-0.2Ti UHP.
Chitose, Hiromasa*; Hotta, Akitoshi*; Onuki, Akira; Tamai, Hidesada; Yoshida, Hiroyuki; Shibata, Mitsuhiko; Akimoto, Hajime; Fujimura, Ken*
no journal, ,
no abstracts in English
Tamai, Hidesada; Onuki, Akira; Shibata, Mitsuhiko; Akimoto, Hajime; Chitose, Hiromasa*; Hotta, Akitoshi*; Fujimura, Ken*
no journal, ,
no abstracts in English
Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
no journal, ,
no abstracts in English
Ishijima, Yasuhiro; Ioka, Ikuo; Kiuchi, Kiyoshi; Usami, Koji; Kato, Yoshiaki; Fujimura, Ken*
no journal, ,
no abstracts in English
Aizawa, Kosuke; Fujimura, Ken; Hirata, Shingo; Kasahara, Naoto
no journal, ,
no abstracts in English
