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Kawasaki, Nobuchika; Hosogai, Hiromi*; Furuhashi, Ichiro*; Kasahara, Naoto
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.1, p.959 - 960, 2006/09
Thermal transient stress at core support structure of advanced fast reactor was evaluated using thermal hydraulic-structure total analysis method with experimental design. Maximum thermal stress is calculated 15
18% larger than nominal thermal stress by uncertainty of system parameters. Maximum thermal stress was evaluated 6368% larger than nominal thermal stress when predicted by the past deign method, therefore about 40% excessive imaginary stress could be appropriate by thermal hydraulic-structure total analysis.
Igari, Toshihide*; Takao, Nobuyuki*; Otani, Tomomi*; Shibamoto, Hiroshi; Kasahara, Naoto
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.1, p.957 - 958, 2006/09
no abstracts in English
Watanabe, Osamu*; Bubphachot, B.*; Kasahara, Naoto; Kawasaki, Nobuchika
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.1, p.117 - 118, 2006/09
This paper is to evaluate the fatigue strength for the perforated plate at elevated temperature based on Stress Redistribution Locus (SRL) method. The test specimens are made of SUS304 stainless steel, and temperature is kept to be 550
C degree, and the geometry of the perforated plate specimens are changed by number of holes and diameter size of the hole. The SRL method is used to predict number of cycle of crack initiation (Nc), and the failure (Nf) for the perforated plate.
Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*; Aoki, Takayuki*; Xu, Z.*
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.2, p.39 - 40, 2006/09
Three-dimensional large-scale numerical simulations were carried out to predict the complicated water-vapor two-phase flow characteristics in a fuel bundle of an advanced light water reactor. Conventional analysis methods with a two-fluid model need composition equations and empirical correlations based on the experimental data. Therefore, it is difficult to obtain high prediction accuracy when experimental data are nothing. Then, a new two-phase flow analysis method was proposed and the TPFIT code was developed. This paper describes the predicted liquid film, bubbly and droplet flow behavior in the simulated fuel channels with the TPFIT code, and the predicted two-phase flow behavior around a curved fuel rod with the FLUENT code which is one of the most famous commercial code. From the present results, the high prospect was acquired on the possibility of development of the thermal design procedure of the advanced nuclear reactors by large-scale simulations.
Shiina, Yasuaki; Ishikawa, Kota*; Hishida, Makoto*; Tanaka, Gaku*
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.173 - 174, 2006/09
no abstracts in English
Muramatsu, Toshiharu
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.7, p.139 - 140, 2006/09
no abstracts in English
Kondo, Masaya
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.7, p.69 - 70, 2006/09
no abstracts in English
Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki
no journal, ,
no abstracts in English
Watanabe, Tadashi
no journal, ,
no abstracts in English
Takata, Takashi*; Yamaguchi, Akira*; Suda, Kazunori; Ohshima, Hiroyuki
no journal, ,
When a heat transfer tube fails in a steam generator of liquid sodium cooled fast reactor, highly pressurized water or steam inside the tube blows down into liquid sodium that exists in the shell side and thus the sodium-water reaction will occur. The reacting zone, in which high temperature appears and corrosive sodium compound such as sodium hydroxide (NaOH) exists, will affect a structural integrity of neighboring heat transfer tubes and shell structure. Hence, an investigation of the reacting zone in the sodium-water reaction phenomena is of great importance for safety evaluation in the steam generator. A numerical method of multi-dimensional and multi-phase thermal hydraulics coupled with the sodium-water reaction has been developed for this purpose. In the present paper, the influence of the initial pressure in the shell side on the reacting zone has been investigated numerically, as well as the geometric configuration of pin-bundles.
Matsuoka, Hiroshi; Kikuchi, Noriko*
no journal, ,
A simulation approach by "Package Flow Model (PFM)" was previously proposed, which enables us to intuitively understand the dynamic behavior of various systems. The model does not directly consider the physical process of given actual system, but will replace it a simple visual mechanism (PFM) which is equivalent only in a view point of "time delay" of the system-response. A total system such as a nuclear reactor fluid system is modeled by a combination of several PFMs. Using the PFMs network, we can instantly calculate many transient phenomena of the system even by a notebook-type personal computer. Experts' intuition or experiences can be enhanced by using it with effective representation methods. In addition, PFM method has a capability to develop into an more effective simulation method of total system by including Lattice Gas Methods applied to its subsystems because the calculation processes are in common, i.e., neural networks.