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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.
Yokoyama, Kenji; Hosogai, Hiromi*; Chiba, Go; Kasahara, Naoto; Ishikawa, Makoto
JNC TN9400 2004-022, 162 Pages, 2004/04
JNC-TN9400-2004-022.pdf:13.47MB
In the fast reactor development, numerical simulation using analysis code plays an important role for complementing theory and experiment. In order to efficiently advance the research and development of fast reactors, JNC promotes the development of next generation simulation code (NGSC). In this report, research result by prototyping which carried out for the conceptual design of the NGSC is described. From the viewpoint of the cooperative research with CEA (Commissariat a l'Energie Atomique) in France, a trend survey on several platforms for numerical analysis and an applicability evaluation of CEA's SALOME platform for the NGSC were carried out. As a result of the evaluation, it is confirmed that SALOME had been satisfied the features of efficiency, openness, universality, expansibility and completeness that are required by the NGSC. In addition, it is confirmed that SALOME had the concept of the control layer required by the NGSC and would be one of the important candidates as a platform of the NGSC. In the field of the structure analysis, the prototype of the PARTS.NET code was reexamined from the viewpoint of class structure and input/output specification in order to improve the data processing efficiency and maintainability. In the field of the reactor physics analysis, a development test of a new code with C++ and a reuse test of an existing code written in Fortran was carried out in view of utilizing SALOME for the NGSC.
Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto
JNC TN9400 2003-021, 205 Pages, 2003/04
JNC-TN9400-2003-021.pdf:8.86MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. Aming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1)Multi-language (SoftWIRE.NET, Visual Basic .NET and Fortran) (2)Fortran90 and (3)Python to make a prototype of the next generation code system. As this result, the followings were comfirmed. (1)It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using visual Basic .NET. (2)The maintenanability of the existing code written by Fortran77 can be improved by using the new features of Fortran90. (3)The toolbox-type code system can be built by using Python.
kasahara, Naoto; Hosogai, Hiromi*; Jimbo, Masakazu*
Transactions of 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT-17) (CD-ROM), 0 Pages, 2003/00
None
Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; ; ; Ishikawa, Makoto
JNC TN9420 2002-004, 309 Pages, 2002/11
JNC-TN9420-2002-004.pdf:11.4MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the phisical propaties and engineering models are replesented as a programming languare or a diagams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language(EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML.
Kasahara, Naoto; Jimbo, Masakazu*; Hosogai, Hiromi*
Saikuru Kiko Giho, (16), p.81 - 92, 2002/09
None
; Hosogai, Hiromi*; Furuhashi, Ichiro*; kasahara, Naoto
JNC TN9400 2001-121, 44 Pages, 2002/02
JNC-TN9400-2001-121.pdf:1.16MB
PARTS, Program for Arbitrary Real Time Simulation is being developed: it is expected to make great contribution to fast reactor components' designing work by enabling integration of thermal hydraulic and structural analysis. Since PARTS is a tool to perform the integrated thermal hydraulic-structural analysis under various conditions, it needs to calculate rapidly. At the point, the Green function method seems to be the most Promising stress analysis procedure for PARTS. The Green function method figures out thermal transient stress arising in structures in the form of convolute integration corresponding to fluids' step temperature changes. It is expected to calculate faster than Finite Elemental Method. Hitherto, the Green function method has been used to describe the response to sole thermal fluid with a constant heat transfer coefficient. In this report, the Green function method is extended to cope with a cylinder touching two thermal fluids with variable heat transfer coefficients (inside and outside surfaces contacting with primary and secondary coolants respectively) and is confirmed to be sufficiently applicable to such condition.
Hosogai, Hiromi*; Kawasaki, Nobuchika; kasahara, Naoto
JNC TN9520 99-002, 106 Pages, 1998/12
None
Hosogai, Hiromi*; Kawasaki, Nobuchika; kasahara, Naoto
JNC TN9520 99-001, 545 Pages, 1998/12
None
Hosogai, Hiromi*; kasahara, Naoto
PNC TN9460 98-002, 240 Pages, 1998/07
Object oriented Program "SONSHO" predicts creep fatigue damage factors based on Elevated Temperature Structural Design Guide for "Monju" and other various procedures from stress classification data obtained from structural analysis results. From view point of program implementation, it is required that external programs interface and frequent revise from update of material data and creep fatigue evaluation methods. 0bject oriented approach was continuously introduced to improve these aspects of the program, Version 4.0 has the following new functions. (1)Material strength library was implemented as an independent program module based on Microsoft Active X control and 32bit DLL technologies, which can be accessed by general Windows programs. (2)Self instruction system "Wizard" enables manual less operation. (3)Microsoft common object model (COM) was adopted for program jnterface, and this program can communicate with Excel sheet data on memory. Sonsho Ver.4.0 can work on Windows 95 or Windows NT 4.0. Microsoft Visual Basic 5.0 (Enterprose Edition) and Microsoft FORTRAN Power Station 4.0 were adopted for program.
