Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Fukuda, Kodai; Yamane, Yuichi
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10
This study presents the results of multiphysics analysis, which investigates the change of reactivity caused by the motion of fluids, of Windscale Works criticality accident. The purpose of this study is to confirm previously reported trends of emulsion formation and increase in reactivity by the multi-physics analysis which takes the motion of fluids into account. Continuous energy Monte Carlo code MVP3 was used to calculate reactivity based on the material distribution obtained by CFD calculation using OpenFOAM. An interface program in python was developed to transfer data from OpenFOAM to MVP3. The change of reactivity caused by the motion of solutions was calculated without considering the generation of heat by fissions in a system that simulated the transfer vessel at Windscale Works. As a result, trends of emulsion formation and increase in reactivity were confirmed. The influence of the resolution of the calculation system on the results was also discussed.
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.
Yamamoto, Masahiro
Keisan Kogaku, 25(3), p.4105 - 4108, 2020/07
Recently, some attempts using Multi-Physics simulation for corrosion problems, especially crevice corrosion, have been increasing. Corrosion undergoes by electrochemical reaction. The numerical calculation procedure is used a non-liner equation. Furthermore, this reaction is affected by environmental factors, i.e. composition, amount and mobility of chemical species and redox potential. These values change with time by corrosion process itself. This report, these needs for Multi-Physics calculations are introduced.
Fukuda, Kodai; Yamane, Yuichi
no journal, ,
A preliminary analysis of the criticality incident at Windscale Works in 1970 was performed. The open-source CFD tool box OpenFOAM and the Monte Carlo neutron transport calculation code MVP3.0 were used for the analysis. As the result, the progress of the incident and the change of reactivity during the incident were confirmed.
