Safety studies of plasma-wall events with AINA code for Japanese DEMO

日本の原型炉におけるAINAコードによるプラズマー壁相互作用の安全解析

Rivas, J. C.*; 中村 誠; 染谷 洋二; 高瀬 治彦; 飛田 健次; Dies, J.*; Blas, A. de*; Fabbri, M.*; Riego, A.*

Rivas, J. C.*; Nakamura, Makoto; Someya, Yoji; Takase, Haruhiko; Tobita, Kenji; Dies, J.*; Blas, A. de*; Fabbri, M.*; Riego, A.*

日欧幅広いアプローチ活動における国際協力研究の1つとして、プラズマ-炉内機器の過渡応答を解析するAINAコードを日本の原型炉案(水冷却ペブルベッドブランケット)向けに改造・適用した。2014年には日本案のブランケットが適用できるように炉内機器モデルを、2015年にはプラズマの物理モデルを改良して冷却材喪失や過出力事象を解析したところ、予備的な検討では安全機器の追加が必要である結果が得られたので、本会議で報告する。

In the frame of JAPAN-EU collaborative work for development of AINA code in 2014-2016, a version of AINA code has been developed for the Japanese DEMO WCPB design. During 2014, the AINA code was adapted from ITER to this new mission. A breeding blanket model was implemented in code. The configuration was changed to implement the design parameters of DEMO reactor. Finally, safety studies of plasma-wall transients affecting blanket region were performed. During 2015, plasma models were improved both for plasma core and for divertor (improved SOL model). Safety analyses affecting divertor were performed, considering thermohydraulic accidents and plasma transients where loss of control function was assumed. First analyses performed for the Japanese DEMO design show the behavior of the reactor during Ex-Vessel LOCA and during overpower events. The preliminary conclusions point to the possibility of considering the plasma control system as a safety important component.