A simulation study of large power handling in the divertor for a Demo reactor

Nobuyuki Asakura, Katsuhiro Shimizu, Kazuo Hoshino, Kenji Tobita, Shinsuke Tokunaga, Tomonori Takizuka

研究成果: Article査読

48 被引用数 (Scopus)

抄録

Power exhaust for a 3 GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500 MW and radiation loss of 460 MW, and a fixed core-edge boundary of 7 × 1019 m-3 were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport and radiation , was reduced from 15-16 MW m-2 (Ne and Ar) to 11 MW m-2 for the higher Z (Kr), and extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak qtarget value was decreased to 12 MW m-2, where neutral heat load became comparable to and due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of χ and D enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both qtarget near the separatrix and Te at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.

本文言語English
論文番号123013
ジャーナルNuclear Fusion
53
12
DOI
出版ステータスPublished - 2013 12月
外部発表はい

ASJC Scopus subject areas

  • 核物理学および高エネルギー物理学
  • 凝縮系物理学

フィンガープリント

「A simulation study of large power handling in the divertor for a Demo reactor」の研究トピックを掘り下げます。これらがまとまってユニークなフィンガープリントを構成します。

引用スタイル