Performance of the LH antenna with carbon grill in JT-60U

Seki, M.; Moriyama, S.; Shinozaki, S.; Hasegawa, Kouichi; Hiranai, S.; Yokokura, K.; Shimono, M.; Terakado, M.; Fujii, T.

doi:10.1016/j.fusengdes.2005.06.002

Cited by 4 publications

(4 citation statements)

References 6 publications

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“…The gap width is optimized with respect to better coupling and less heat load to the antenna from bombardment. A carbon grill [29] installed in front of the antenna was removed in this campaign.…”

Section: Current Drivers In Jt-60umentioning

confidence: 99%

Off-axis current drive and real-time control of current profile in JT-60U

et al. 2008

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Aiming at optimization of current profile in high-β plasmas for higher confinement and stability, a real-time control system of the minimum of the safety factor (q min) using the off-axis current drive has been developed. The off-axis current drive can raise the safety factor in the centre and help to avoid instability that limits the performance of the plasma. The system controls the injection power of lower-hybrid waves, and hence its off-axis driven current in order to control q min. The real-time control of q min is demonstrated in a high-β plasma, where q min follows the temporally changing reference q min,ref from 1.3 to 1.7. Applying the control to another high-β discharge (βN = 1.7, βp = 1.5) with m/n = 2/1 neo-classical tearing mode (NTM), q min was raised above 2 and the NTM was suppressed. The stored energy increased by 16% with the NTM suppressed, since the resonant rational surface was eliminated. For the future use for current profile control, current density profile for off-axis neutral beam current drive (NBCD) is for the first time measured, using the motional Stark effect diagnostic. Spatially localized NBCD profile was clearly observed at the normalized minor radius ρ of about 0.6–0.8. The location was also confirmed by multi-chordal neutron emission profile measurement. The total amount of the measured beam driven current was consistent with the theoretical calculation using the ACCOME code. The CD location in the calculation was inward shifted than the measurement.

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Section: Current Drivers In Jt-60umentioning

confidence: 99%

Off-axis current drive and real-time control of current profile in JT-60U

et al. 2008

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“…Fabrication aspects. LH launchers have been made with different waveguide counts, array structures, designs, construction techniques and materials: from 2 waveguides [9] up to 144 in a 12 × 12 array [136] in standard grills; from four waveguides [5] up to 384 in a 12 × 32 array [72,73] in multijunctions; using simple extruded waveguides in grills, but mostly machined from solid blocks (mainly stainless steel, but also titanium [137], copper and zirconium [144], and carbon fibre composite [151]), as well as from plates of various materials (e.g. copper + stainless steel + copper) linked by explosive diffusion [87,143]; assembled by stacking and bolting plates [137,142], by vacuum brazing pre-machined plates [143,146], by electron beam welding and laser beam techniques [143]; most rely on inertial cooling but some use water cooling [87,143,144].…”

Section: The Multijunctionmentioning

confidence: 99%

An ITER-relevant passive active multijunction launcher for lower hybrid current drive in JET-grade plasmas

et al. 2011

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Lower hybrid current drive (LHCD) has been demonstrated to be an important actuator for controlling the current profiles of advanced tokamak plasmas, either in fully non-inductive or in hybrid regimes. To continue such scenario development in JET, and most notably towards extrapolations for ITER, an upgrade of its LH launcher is required to deliver more power (>5 MW) and to deal with a broad array of density situations at the plasma edge more consistently. These may range from steep gradients to ELM activity; extend from as low as the cut-off density nec to large and fast density variations and give rise to strong and rapid reflections of RF power. The passive active multijunction (PAM), advocated as the LH launcher concept for ITER, is claimed to be more resilient to such conditions, run with increased efficiency near nec and even with a small vacuum gap, being also specially appropriate for long pulse operation—all conditions hampering traditional designs. In addition, some of this leverage has been confirmed experimentally at the ITER-relevant FTU with the first PAM ever deployed, and more recently with Tore Supra's PAM. The JET version addressed in this study, with the latter focusing on the physics that drives and supports the adopted technical solutions, allows the injection of up to 5.2 MW at 3.7 GHz and N||peak = 1.9. With directivities better than 70%, reflection coefficients below 2% even for ∼nec and power densities comparable to those foreseen for ITER at 5 GHz, it would validate this concept in ITER-relevant scenarios.

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“…The target performance after the modification are shown in Table 1. To attain these numbers, we have been solving technological issues relating to the thermal limitations for the following heating systems available in JT-60U: The positiveand negative-ion-based NBI (P/N-NBI) [2], the electron cyclotron wave heating (ECH) system [3], and the lower hybrid wave heating (LHH) system [4]. These heating devices were employed to produce advanced plasma operation during the last experimental campaign of Dec.…”

Section: Introductionmentioning

confidence: 99%

“…In the RF systems, lower hybrid (LH) and electron cyclotron (EC) range of frequency, the possible maximum duration has been extended up to 60 s. In the LH heating (LHH) system, a grill mouth made of carbon was developed and bolted to the original stainless steel grill to allow long pulse operation [4]. The achieved pulse length is 18 s with -0.9 MW.…”

Section: Introductionmentioning

confidence: 99%

Current Status of Experimental Study and Device Modifications in JT-60U

Kurihara¹

2005

21st IEEE/NPS Symposium on Fusion Engineering SOFE 05

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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 quasisteady 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.

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Performance of the LH antenna with carbon grill in JT-60U

Cited by 4 publications

References 6 publications

Off-axis current drive and real-time control of current profile in JT-60U

Off-axis current drive and real-time control of current profile in JT-60U

An ITER-relevant passive active multijunction launcher for lower hybrid current drive in JET-grade plasmas

Current Status of Experimental Study and Device Modifications in JT-60U

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