Numerical analysis of tungsten erosion at the plasma facing surface in DEMO reactor

核融合原型炉におけるプラズマ対向壁のタングステン損耗に関する数値解析

星野 一生; 藤間 光徳*; 清水 勝宏; 朝倉 伸幸; 飛田 健次; 畑山 明聖*; 滝塚 知典*

Hoshino, Kazuo; Toma, Mitsunori*; Shimizu, Katsuhiro; Asakura, Nobuyuki; Tobita, Kenji; Hatayama, Akiyoshi*; Takizuka, Tomonori*

Tungsten (W) is the most preferable candidate for the plasma facing material in the DEMO fusion reactor. Although one of advantages of the W armor is the low sputtering yield, sputtering by impurity species seeded for the divertor radiation cooling cannot be ignored. In this study, erosion rate is evaluated by using the IMPGYRO-EDDY code. The background plasma profile was calculated by the integrated divertor code SONIC. Erosion rate significantly changes along the divertor target under the partially detached plasma condition, where the ion and electron temperature near the outer strike point ( 5 cm) is about 1 eV. As long as the detached divertor plasma is sustained, the surface erosion is low even under the condition of significant Ar recycling flux. On the other hand, in the outer SOL region ( 5 cm), the ion temperature is increased to high ( 100 eV) due to low recycling. Therefore, the self-sputtering yield may exceed 1. Erosion rate at the first wall surface is also evaluated.