Compact DEMO, SlimCS: design progress and issues

被引:139
作者
Tobita, K. [1 ]
Nishio, S. [1 ]
Enoeda, M. [1 ]
Kawashima, H. [1 ]
Kurita, G. [1 ]
Tanigawa, H. [1 ]
Nakamura, H. [1 ]
Honda, M. [1 ]
Saito, A. [1 ]
Sato, S. [1 ]
Hayashi, T. [1 ]
Asakura, N. [1 ]
Sakurai, S. [1 ]
Nishitani, T. [1 ]
Ozeki, T. [1 ]
Ando, M. [1 ]
Ezato, K. [1 ]
Hamamatsu, K. [1 ]
Hirose, T. [1 ]
Hoshino, T. [1 ]
Ide, S. [1 ]
Inoue, T. [1 ]
Isono, T. [1 ]
Liu, C. [1 ]
Kakudate, S. [1 ]
Kawamura, Y. [1 ]
Mori, S. [1 ]
Nakamichi, M. [1 ]
Nishi, H. [1 ]
Nozawa, T. [1 ]
Ochiai, K. [1 ]
Ogiwara, H. [1 ]
Oyama, N. [1 ]
Sakamoto, K. [1 ]
Sakamoto, Y. [1 ]
Seki, Y. [1 ]
Shibama, Y. [1 ]
Shimizu, K. [1 ]
Suzuki, S. [1 ]
Takahashi, K. [1 ]
Tanigawa, H. [1 ]
Tsuru, D. [1 ]
Yamanishi, T. [1 ]
Yoshida, T. [1 ]
机构
[1] Japan Atom Energy Agcy, Naka, Ibaraki 3110193, Japan
关键词
THERMAL-CONDUCTIVITY; COMPONENTS; PLANT;
D O I
10.1088/0029-5515/49/7/075029
中图分类号
O35 [流体力学]; O53 [等离子体物理学];
学科分类号
070204 ; 080103 ; 080704 ;
摘要
The design progress in a compact low aspect ratio (low A) DEMO reactor, 'SlimCS', and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m(-2) or lower, which can be a critical constraint for determining a handling power of DEMO.
引用
收藏
页数:10
相关论文
共 24 条
[1]   Maximum implementation capacity of fusion power reactors [J].
Asaoka, Y ;
Okano, K ;
Yoshida, T ;
Hiwatari, R ;
Tokimatsu, K .
FUSION TECHNOLOGY, 2001, 39 (02) :518-522
[2]  
*AT EN COMM ADV CO, 2005, NAT POL FUT NUCL FUS
[3]   ARIES-RS magnet systems [J].
Bromberg, L ;
Titus, P ;
Schultz, JS ;
Sidorov, M ;
Pourrahimi, S .
FUSION ENGINEERING AND DESIGN, 1997, 38 (1-2) :159-188
[4]  
Enoeda M, 2003, NUCL FUSION, V43, P1837, DOI 10.1088/0029-5515/43/12/026
[5]   Dynamic transport simulation code including plasma rotation and radial electric field [J].
Honda, M. ;
Fukuyama, A. .
JOURNAL OF COMPUTATIONAL PHYSICS, 2008, 227 (05) :2808-2844
[6]   ITER TF model coil assembly, commissioning and instrumentation [J].
Hurd, FH ;
Fillunger, H ;
Zahn, G ;
Ulbricht, A ;
Libeyre, P ;
Theisen, E ;
Beaudet, F .
FUSION ENGINEERING AND DESIGN, 2001, 58-59 :171-176
[7]   Development of an extensive database of mechanical and physical properties for reduced-activation martensitic steel F82H [J].
Jitsukawa, S ;
Tamura, M ;
van der Schaaf, B ;
Klueh, RL ;
Alamo, A ;
Petersen, C ;
Schirra, M ;
Spaetig, P ;
Odette, GR ;
Tavassoli, AA ;
Shiba, K ;
Kohyama, A ;
Kimura, A .
JOURNAL OF NUCLEAR MATERIALS, 2002, 307 (1 SUPPL.) :179-186
[8]   Development of integrated SOL/divertor code and simulation study of the JT-60U/JT-60SA tokamaks [J].
Kawashima, H. ;
Shimizu, K. ;
Takizuka, T. .
PLASMA PHYSICS AND CONTROLLED FUSION, 2007, 49 (07) :S77-S85
[9]   Simulation study for divertor design to handle huge exhaust power in the SlimCS DEMO reactor [J].
Kawashima, H. ;
Shimizu, K. ;
Takizuka, T. ;
Tobita, K. ;
Nishio, S. ;
Sakurai, S. ;
Takenaga, H. .
NUCLEAR FUSION, 2009, 49 (06)
[10]   DEMO plant design beyond ITER [J].
Konishi, S ;
Nishio, S ;
Tobita, K .
FUSION ENGINEERING AND DESIGN, 2002, 63-64 :11-17