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Matsuba, Kenichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.13 - 14, 2007/06
no abstracts in English
Takamatsu, Kuniyoshi; Takeda, Tetsuaki; Nakagawa, Shigeaki; Goto, Minoru
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.213 - 214, 2007/06
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features and to improve the safety technology and design methodology of high temperature gas-cooled reactors (HTGRs). The numerical analysis code was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel flow channels and twenty temperature coefficients in the core. This paper describes an analytical result of the loss of partial coolant flow test using the newly developed code. The analytical result of transient reactor power shows good agreement with the measured value during the test. Moreover, this paper refers to an analytical result of the loss of coolant flow test. The reactor power decreases to decay heat level due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again later, the peak power value is very small.
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.
Liu, W.; Kureta, Masatoshi; Yoshida, Hiroyuki; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.103 - 106, 2007/06
With using the research achievements on thermal-hydraulic characteristics so far derived for the target tight lattice core, this paper studies the thermal feasibility of the designed 1356 MWe FLWR core using a modified transient analysis code TRAC-BF1. The newest critical power correlation developed at JAEA for tight lattice rod bundles is implemented to judge the occurrence of boiling transition from nucleate boiling to film boiling. The pressure drop in two-phase flow region is evaluated by Martinelli-Nelson two-phase multiplier. In the analyses, the 900 fuel assemblies in the designed core are modeled into 12 fuel channels according to the relative mass and power distributions. Analyses to the postulated abnormal transient events that may be possibly met in the operation of the FLWR are performed with 
MCPRs being evaluated. The necessary coolant flow rate then is calculated based on the evaluated MCPRs. As the results, for a natural circulation type FLWR, the operation limited MCPR is 1.19. For a forced one, it is 1.32.
Iwatsuki, Jin; Terada, Atsuhiko; Noguchi, Hiroki; Ishikura, Shuichi; Takahashi, Toshio*; Hino, Ryutaro
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.285 - 286, 2007/06
The Japan Atomic Energy Agency has been proceeding design works on a pilot test plant of 30Nm
/hr-scale hydrogen production by using thermo-chemical water splitting process by iodine and sulfur (IS process) to contribute future hydrogen economy using high-temperature gas-cooled reactors. In parallel to the design works, key engineering issues on corrosion-resistant pipelines and its sealing using dish-type springs were examined under high temperature conditions. This paper introduces experimental results of heat cycle test results of a glass-lining pipe and seal performance using dish-type springs which works as thermal expansion absorber of bolts.
Okawachi, Yasushi; Sugino, Kazuteru; Sekine, Takashi; Soga, Tomonori; Kitamura, Ryoichi; Aoyama, Takafumi
no journal, ,
A mixed oxide containing minor actinides fuel irradiation program is being conducted using the experimental fast reactor Joyo to develop a low decontamination TRU fuel cycle technology. In the program, irradiation test subassemblies were fabricated, which contains test fuel pins including MOX fuel containing 5% americium in maximum and MOX fuel containing 2% americium and 2% neptunium. Short-term and steady-state irradiation experiments were planned in Joyo. The short-term test of 10 minutes irradiation was conducted in May 2006 to research early thermal behavior of MA-MOX fuel. Thus, the operation fulfilling the necessary test conditions has been achieved.
Watanabe, Hironori; Mitsutake, Toru*; Kakizaki, Sadayuki*; Takase, Kazuyuki
no journal, ,
The electro-void fraction meter (Capacitance Type meter) is practical for high void fraction measurement. It can be used with various shapes of flow conduits such as round, rectangular and rod-bundle geometries. The principle of the meter is that the electrical capacitance of a gas-liquid two-phase flow changes with respect to the void fraction. High-frequency power supply enables to measure the void fraction of the pure water. It was confirmed by an air-water two-phase flow experiment that void fraction can be obtained in real time by measuring the capacitance of the two-phase flow. Void fraction ranging from 0 to more than 0.9 in a 37-rod bundle was successfully measured under 7MPa pressure conditions.
Aizawa, Kosuke; Kisohara, Naoyuki; Kotake, Shoji; Sherwood, D. V.*
no journal, ,
no abstracts in English
Misawa, Takeharu; Onuki, Akira; Katsuyama, Kozo; Nakamura, Yasuo; Akimoto, Hajime
no journal, ,
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime
no journal, ,
Uchibori, Akihiro; Ohshima, Hiroyuki; Takata, Takashi*; Yamaguchi, Akira*
no journal, ,
A numerical method for the multi-phase multi-component flows with sodium-water reaction was developed. When the pressurized water leaks from a failed heat transfer tube in the steam generator of sodium cooled fast breeder reactors, the water will react with the sodium outside the tubes. In order to simulate the multi-phase multi-component flows, the multi-fluid model and the advection-diffusion equations are used in our numerical method. Production rates of the components such as the sodium hydroxide and the hydrogen by sodium-water reaction are calculated by using the surface reaction model and the gas phase reaction model. Through the numerical analysis of the sodium-water reaction test SWAT-1R, validity of the numerical method was demonstrated.
Ito, Kei; Ohshima, Hiroyuki
no journal, ,
Gas Entrainment (GE) phenomena in the basic experimental apparatus were studied numerically to make clear the dominant parameters on the GE phenomena that might occur in FBRs. In the numerical simulations, the PLIC type VOF model was employed to calculate these complicated free surface behaviors. The result of numerical simulation under the experimental conditions was compared to the experimental results to evaluate the accuracy of the numerical method. In addition, sensitivities of the suction velocity and the geometric configuration around the suction mouth on the GE behaviors were investigated. As a result, the numerical simulation showed good agreement with experimental data. In addition, two types of the GE were observed in this simulation result. The geometric configuration around the suction mouth also affected the GE behavior. These simulation results implied that the numerical simulation has enough potential to be used for the GE investigation.