Refine your search:     
Report No.
Search Results: Records 1-5 displayed on this page of 5
  • 1

Presentation/Publication Type

Initialising ...


Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...


Initialising ...


Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Development of fast reactor structural integrity monitoring technology using optical fiber sensors

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

Journal Articles

Development of analytical methodology regarding reactor performance and safety characteristics of HTGR; Loss of coolant flow tests

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.

Journal Articles

Study on cross flow phenomena in a tight-lattice rod bundle by statistical method

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.

Journal Articles

A Study on the thermal feasibility of 1356 MWe innovative water reactor for flexible fuel cycle (FLWR)

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 $$Delta$$MCPRs being evaluated. The necessary coolant flow rate then is calculated based on the evaluated $$Delta$$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.

Journal Articles

Development of chemical reactors for thermo-chemical water-splitting IS process

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$$^{3}$$/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.

5 (Records 1-5 displayed on this page)
  • 1