DocumentCode :
2758424
Title :
Current Status of Experimental Study and Device Modifications in JT-60U
Author :
Kurihara, K.
Author_Institution :
Naka Fusion Res. Establ., Japan Atomic Energy Res. Inst., Ibaraki
fYear :
2005
fDate :
Sept. 2005
Firstpage :
1
Lastpage :
6
Abstract :
Since tokamak magnetic fusion research has just made a step forward to an international collaborative project ITER, the existing tokamaks including JT-60 are expected to explore more advanced operation scenarios. To test those scenarios in the JT-60 experiment, the discharge pulse length and the duration time of additional NBI/RF heating were extended to 65 s and 30 s/60 s, respectively, in 2003 with modification of the corresponding control systems for power supplies and heating devices. The experimental campaign in 2003-2004 after the above modifications has ended up with the following significant results: (a) The high bootstrap current ratio of 75 % was sustained for 7.4 s in an R/S plasma, (b) Normalized beta value of 2.3 was also done for 22.3 s in a high-beta H-mode plasma, (c) The quasi-steady state beta value was increased to 3.0 with a pulse of 6.2 s with NTM suppression by ECCD. For further exploration toward high performance plasmas, the following modifications will be or has been conducted: (1) To minimize the power loss from a plasma at the region of toroidal field ripple, the 8Cr ferritic steel tiles, having a similar magnetic property to the low activation ferritic material for a DEMO reactor, are being equipped on the first wall of the JT-60 vacuum vessel. (2) Since plasma shape and current profile in the poloidal cross-section are expected to be reproduced in real time to optimize a plasma performance with suppressing the plasma instabilities, a precise reproduction method has been installed in the plasma control system. In this report, the current status of plasma experimental study will be presented together with on-going device modifications in JT-60U
Keywords :
Tokamak devices; fusion reactor divertors; plasma hybrid waves; plasma instability; plasma magnetohydrodynamics; plasma radiofrequency heating; plasma toroidal confinement; 8Cr ferritic steel tiles; DEMO reactor; ECCD; JT-60 vacuum vessel; MHD stability; NTM suppression; R/S plasma; W-shaped divertor; additional NBI/RF heating; advanced operation scenarios; discharge pulse length; high bootstrap current ratio; high performance plasmas; high-beta H-mode plasma; international collaborative project ITER; low activation ferritic material; lower hybrid wave heating; magnetic property; on-going device modifications; plasma control system; plasma current profile; plasma instabilities; plasma shape control; poloidal cross-section; power supplies; quasisteady state beta value; tokamak magnetic fusion research; toroidal field ripple; Control systems; Heating; Magnetic materials; Performance loss; Plasma devices; Plasma materials processing; Plasma properties; Pulsed power supplies; Shape control; Tokamaks; JT-60U; advanced operation; device modifications; ferritic steel; plasma current profile; plasma experiment; plasma shape control; tokamak;
fLanguage :
English
Publisher :
ieee
Conference_Titel :
Fusion Engineering 2005, Twenty-First IEEE/NPS Symposium on
Conference_Location :
Knoxville, TN
Print_ISBN :
0-4244-0150-X
Electronic_ISBN :
0-4244-0150-X
Type :
conf
DOI :
10.1109/FUSION.2005.252846
Filename :
4018880
Link To Document :
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