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Aizawa, Kosuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.229 - 230, 2010/06
A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the "Fast Reactor Cycle Technology Development (FaCT)" project in Japan. JSFR adopts a selector-valve mechanism for the Failed Fuel Detection and Location (FFDL) system. The selector-valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an Upper Internal Structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for identifying under-slit failed fuel subassemblies has been demonstrated by water experiments.
Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.287 - 288, 2010/06
no abstracts in English
Kato, Atsushi; Negishi, Kazuo; Akiyama, Yo*; Kubo, Shigenobu*
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.231 - 232, 2010/06
Reduction of plant construction cost is one of the most important issues for commercialization of fast reactors. From this point of view, an innovative containment vessel adopting steel plate reinforced concrete structure (SCCV) is developed for Japan Sodium-cooled Fast Reactor (JSFR). Although SC structure is generally in practical use, performance after exposing high temperature is not investigated. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the design feature, the design and evaluation conditions for SCCV of JSFR as well as the construction method are summarized.
Tanaka, Masaaki; Kurokawa, Koji*; Takita, Hiroki*; Monji, Hideaki*; Ohshima, Hiroyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.221 - 224, 2010/06
Numerical simulations for a T-pipe consisting of a rectangular duct for main stream and a circular pipe for branch stream were performed to investigate relation between large-scale eddy structure formation in the mixing area and temperature fluctuation generation on the wall. Water experiment and numerical simulation in the T-pipe to investigate fluid-structure thermal interaction were conducted. The numerical results indicate that the fluid-structure thermal interaction is necessarily considered for thermal fatigue estimation in the thermal striping phenomena.
Yoshida, Hiroyuki; Hosoi, Hideaki; Suzuki, Takayuki*; Takase, Kazuyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.273 - 276, 2010/06
no abstracts in English
Hosoi, Hideaki*; Yoshida, Hiroyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.277 - 278, 2010/06
no abstracts in English
Nakatsuka, Toru; Liu, W.; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.279 - 280, 2010/06
To assess the stability of once-through steam generators in FBR, Japan Atomic Energy Agency has been developing a prediction method for thermal-hydraulic instability based on system analysis code TRAC-BF1. In the present paper, to simulate the primary coolant in steam generators, thermal property of sodium was incorporated to the code and the VESSEL component was improved to handle two different fluids of primary sodium and secondary water. These added functions were assessed with a simplified steam generator model calculation by altering primary coolant fluid as water and sodium. It was confirmed that heat transfer at steam generators was properly evaluated for the case that primary coolant is sodium as well as water.