平成26年度合同会合2014.8.1筑波大学 大学会館モデリングの現状トカマクダイバータシミュレーション星野一生原子力機構contents 1.国際的進展(PSI2014概要) 2.国内活動(原型炉ダイバータシミュレーション) 3.日欧共同研究1.国際的進展（PSI2014概要） http://psi2014.nifs.ac.jp/talks.html を参照してください。 2.国内活動(原型炉ダイバータシミュレーション)
● Enhancement of P rad div and P rad SOL+edge and “Full detachment” were required. ⇒ Divertor operation consistent with qtarget (incl. radiation load) < 10MWm -² was seen only in specific conditions such as enhanced diffusion. • Seeding impurity selection: higher Z (Ar/Kr) is preferable to increase Prad issue: dilution in core plasma, and potentially, Pragedge restriction for good confinement. • Longer leg divertor design: effective for full detachment and Praddiv enhancement issues: reduction in ion and neutral fluxes, and optimization with He exhaust • Plasma diffusion – large impact on detachment and energy dissipation, suggesting that global/ local enhancement promotes full detachment. Issues: extrapolation to Demo, techniques. [1] K. Hoshino, et al. Contrib. Plasma Phys. 52 (2012) 550.  [2] N. Asakura, et al. Nucl. Fusion, 53 (2013) 123013.
•  Design parameters for power exhaust with Pout/R = 30-60: 2-4 times larger than ITER P fus = 1.5-2GW level (P out = 300-400MW) and R = 7-9 m. • Large f rad = 0.8-0.9 and ITER divertor technology (W&Cu-alloy) will be still required, while experiments with high H-factor and radiation loss are limited up to frad = 0.7-0.8. ⇒ Developments of the divertor physics and engineering/technology are necessary. Simple estimation of peak heat load: Demo design     n sol&div, Imp. seeding Transport & Geometry design Detachment Divertor physics design study will decrease Atarget in Figure.
Pout = 500 ⇒ 320MW (P fus decreases from 3GW to 2 GW) SONIC solution of f rad ~ 92% (P rad =295MW) was obtained by Ar puff of 10 Pa/m³s Impurity radiation moves upstream ⇒ detached region further extends to 12cm. ⇒ The peak heat load decreases to 6MW/m². MW/m³ Note: thermal instability at the divertor will be investigated in future.
Radiation region moves near target with reduction from f rad = 92% (P rad tot = 295MW) to 80% (256MW), where nearly full-detached plasma becomes partially-detached. ⇒ Peak heat load increases from 6 to 10 MW/m². Lower frad < 80% seems to be difficult in the SlimCS-size divertor. Studies in 2014: • Impurity (Ne/Ar/Kr) seeding and long-leg option will be compared to P fus = 3GW. • Divertor operation in similar geometry for New concept design (R n ~ 8m)
SONIC simulation of the divertor plasma in the new Demo design with reduced Pfus: large radiation loss case (f rad =92%) showed that full detached was enhanced. ➡  thermal instability of the divertor plasma occurs. Calculations of lower radiation cases (f rad = 70-85%) are now in progress. Input parameters are the same At core boundary r/a~0.95: Exhausted power from core Pout is given nD+=7.0x1019 m-3 Transport coeff. (same as ITER calc.): D=0.3m²/s, c=1.0m²/s D gas puff: G puff=0.8x10²³ s-¹  (div. + sol) pumping speed:  Spump=200 m³/s 3.日欧共同研究
Comparison DSMC and NEUT2D for subdivertor gas flows in JT60-SA Kick-off meeting at JET on 22/3/2014 C. Gleason-González, S. Varoutis, K. Hoshino, H. Utoh , N. Asakura, T. Nakano Boundary conditions defined and given: 2nd meeting (remote) on 8/5/2014 C. Gleason-González, S. Varoutis, K. Hoshino, K. Shimizu, S.Sakurai, N. Asakura, T. Nakano 1st output from DSMC code: Some difference found 🌸 The boundary conditions are being checked 🌸 Simplified calculation model was suggested
デタッチメントモデリングの改善を目標に、今年からコード間ベンチマークを再開(cf. ITPA DSOL mtg. 2009 )すでにSOLPSやEDGE2Dで解析の進んでいるJET-ILWを対象に、 SONICシミュレーションに着手初期結果について、10月のITPA (DSOL)で報告予定
1.国際的進展
実験データを補完的に用いた解析が精力的に進められている。実験データの再現性は高くなってきているが、unconventional assumptionによるところが大きいと思われ、今後モデリングが必要。
2.国内活動
原型炉ダイバータとしては、原型炉仕様の再検討を受け、パラメータサーベイを開始。課題はあるものの2GW Prad/Pout=92%で現状の工学設計と取り合いができる可能性が見えてきた。 JT-60SAの解析については、燃料・不純物ガスの組み合わせを変えたようなパラメータサーベイが進行中。またW壁の解析も始められている。
3.日欧共同研究
サブダイバータの中性粒子輸送、デタッチメントモデル改善に向けたコード間ベンチマークが進行中