国際会議・海外発表
| 日付 | 講演・会議・掲載誌名など | テーマ |
|---|---|---|
| 2024年11月10日～13日 | NTHAS13 | Development of Numerical Evaluation Method for Heat Transportation with Sodium Mist in the Cover gas Region of Sodium-Cooled Fast Reactor |
| Code-to-code Comparison Through Benchmark Analyses of Run-E1 Sodium Spray Fire Test | ||
| 2024年10月7日～11日 | PSAM17&ASRAM2024 | Benchmark on Dynamic PRA with Simplified Decay Heat Removal System Model of Sodium Fast Reactor– Part1 (Benchmark Analysis Condition and Thermodynamic Model Results) |
| Benchmark on Dynamic PRA with Simplified Decay Heat Removal System Model of Sodium Fast Reactor- Part2 (DPRA Problem Description and Benchmark Results) | ||
| 2024年10月6日～10日 | GLOBAL2024 | Major outcomes of the Franco-Japanese collaboration on R&D for SFR thermal-hydraulics simulation |
| France-Japan Collaboration on Severe Accident Studies in Sodium–Cooled Fast Reactors (2)Methodologies and Calculations of Severe Accident Phases | ||
| Ten Years of Japanese & French Research and Industry Collaboration on GEN-IV-SFR Developments: Outcomes and Prospects | ||
| France-Japan Collaborative Development of Verification, Validation and Uncertainty Quantification for Sodium-Cooled Fast Reactor Neutronics and Shielding Analysis Methods | ||
| 2024年8月25日～28日 | NUTHOS-14 | Application of the GIF Safety Design Criteria and Safety Design Guidelines on Natural Circulation Capability to Next Generation Sodium-Cooled Fast Reactor in Japan |
| 2024年7月28日～8月2日 | ASME PVP2024 | Study on Vertical Sloshing Load Acting on Roof of Cylindrical Tanks in Seismic Wave Excitation |
| Research and Development of Three-Dimensional Isolation System for sodium Cooled Fast Reactor Part7, 8, 9 | ||
| Development of the Buckling Evaluation Method for Large Scale Vessels in Fast Reactors Made of Grade 91 Steel and Austenitic Stainless Steel with Large Initial Imperfections | ||
| 2024年6月9日～12日 | International Congress on Advances in Nuclear Power Plants（ICAPP2024） | Application of the GIF Safety Design Criteria and Safety Design Guidelines on Reactor Shutdown System to Next Generation Sodium-Cooled Fast Reactor in Japan |
| Application of the GIF Safety Design Criteria and Safety Design Guidelines on Decay Heat Removal System to Next Generation Sodium-Cooled Fast Reactor in Japan |