Refine your search:     
Report No.
 - 
Search Results: Records 1-8 displayed on this page of 8
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Evaluation of scale effects in tight-lattice bundles using subchannel analysis

Tamai, Hidesada; Yoshida, Hiroyuki; Masuko, Kenji*; Akimoto, Hajime

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.230 - 236, 2004/12

no abstracts in English

Journal Articles

Numerical simulation of single bubble behavior in rod bundle with interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Tamai, Hidesada; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.264 - 269, 2004/12

no abstracts in English

Journal Articles

Development of a large-scale numerical simulation method on water-vapor two-phase flow through light-water reactor cores

Yoshida, Hiroyuki; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.270 - 276, 2004/12

no abstracts in English

Journal Articles

Planning outline of CHF experiment for small diameter tube in reactor multiple irradiation environment performed in JMTR

Shibamoto, Yasuteru; Yonomoto, Taisuke; Nakamura, Hideo; Nishikizawa, Tomotoshi

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.210 - 214, 2004/11

no abstracts in English

Journal Articles

Numerical Simulation of Melting/Solidification Phenomena Using Extended Finite Element Method

Uchibori, Akihiro; Ohshima, Hiroyuki; Yamaguchi, Akira

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.51 - 55, 2004/11

To evaluate the feasibility of several candidate techniques for a nuclear fuel cycle process, a numerical simulation program using an eXtended Finite Element Method (X-FEM) was developed. The X-FEM is an excellent numerical simulation method for a melting/solidification phenomenon which is a moving boundary problem. The validity of the numerical simulation program was demonstrated through the analysis of an one-dimensional melting/solidification problem.

Journal Articles

Experimental study on gas entrainment at free surface in reactor vessel for a compact sodium cooled fast reactor

Kamide, Hideki; Kobayashi, Jun; Tobita, Akira; Hayashi, Kenji; Kimura, Nobuyuki

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), 179 Pages, 2004/00

In order to realize a fast reactor with advanced economical efficiency, various typed fast reactors have been studied for a several years. As for sodium cooled fast reactor, adoption of a simplified and compact reactor vessel (R/V) with free surface is investigated. The compact R/V may cause gas entrainment phenomena due to increase of velocity in the R/V. Two types of water experiments have been performed in order to prevent the gas entrainment. One is an 1/10th scaled upper plenum model of the R/V. This model is used to optimize flow in whole upper plenum, and the other one is an 1/1.8th scaled partial model of the upper plenum. The partial model experiment evaluates a scale effect on the gas entrainment and mechanism of phenomena. The boundary condition of the partial model is extrapolated from the measured results of the 1/10th scaled full sector model, where flow velocity toward the free surface was reduced by dipped plates. No gas entrainment was observed in the 1/1.8th partial model under the velocity similarity condition as in the designed reactor. Every vortex did not evolve into the gas entrainment though some vortices were found at free surface and in submerged region. It is possible to suppress gas entrainment in this design of the compact R/V.

Journal Articles

VOF Simulation for Gas-Liquid Interface Deformation due to Free Surface Vortex

Ito, Kei; Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.155 - 162, 2004/00

Gas entrainment due to free surface vortex is one of important phenomena to be solved for its onset condition to design the innovative fast breeder reactor because the reactor has the high coolant velocity. The free surface vortex must be studied in detail by both experiments and numerical simulations to make clear the complicated behavior of the vortex gas entrainment. However, an applicability of a numerical simulation to the free surface vortex has not been clear yet. In this paper, numerical simulation for the free surface vortex was conducted to study the sensitivity of some parameters on the simulation accuracy. Moreover, the effect of the horizontal mesh size on the simulation accuracy is discussed because the effect is important to determine the mesh size for the accurate simulation. As a result of the simulation, it became clear that the horizontal mesh size near the vortex center has a significant sensitivity on the simulation accuracy near the vortex center rather than the vertical mesh size. Agreement of the simulation result with the experimental one is highly improved with the decrease of the mesh size. An equation for evaluating simulation accuracy was derived from the extension vortex theory. The relationship between the mesh size and the simulation accuracy was quantified with this equation. The equation showed good agreement with the simulation results in terms of the effect of the horizontal mesh size on the simulation accuracy and was verified to have capability to evaluate the simulation accuracy.

Journal Articles

Numerical Analysis of Gas Core Length Prediction in a Steady Free Surface Vortex

Sakai, Takaaki; Eguchi, Yuzuru*; Iwasaki, Takashi*; Ohshima, Hiroyuki; Yamaguchi, Akira

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.171 - 178, 2004/00

Design method to prevent the gas entrainment from a steady vortex in a cylinder is discussed. The gas core length of the surface vortex dimple is used as a measure of the criterion of the gas entrainment. Burgers vortex model is applied compensationally to predict the gas core length with the practical CFD method. Practical CFD method that did not consider the liquid surface deformation was validated. Macroscopic flow parameters predicted by the CFD method showed reasonable agreements with the measured data in a cylindrical tank. In addition, predicted gas core lengths by using the CFD results showed conservatively longer length than the measured data. The criterion of the gas core length to prevent the gas entrainment was tried to evaluate for the Baum data in order to see the prospects of the design method. In conclusion, it may say that the CFD method has a possibility to evaluate the gas entrainment from a surface vortex.

8 (Records 1-8 displayed on this page)
  • 1