Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

N. Asakura, K. Hoshino, S. Suzuki, S. Tokunaga, Y. Someya, H. Utoh, H. Kudo, Y. Sakamoto, R. Hiwatari, K. Tobita, K. Shimizu, K. Ezato, Y. Seki, N. Ohno, Y. Ueda

研究成果: Article査読

28 被引用数 (Scopus)

抄録

Power exhaust to the divertor and the conceptual design have been investigated for a steadystate DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of Psep = 205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of Pout= 250 MW and the total radiation fraction at the edge, SOL and divertor (Prad/ Pout = 0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load (qtarget) at the attached region was reduced to∼5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak qtarget was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak qtarget of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

本文言語English
論文番号126050
ジャーナルNuclear Fusion
57
12
DOI
出版ステータスPublished - 2017 10 13
外部発表はい

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Condensed Matter Physics

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