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Nakano, Tomohide; Shumack, A.*; Maggi, C. F.*; Reinke, M.*; Lawson, K.*; Coffey, I.*; P
tterich, T.*; Brezinsek, S.*; Lipschultz, B.*; Matthews, G. F.*; et al.
Journal of Physics B; Atomic, Molecular and Optical Physics, 48(14), p.144023_1 - 144023_11, 2015/07
Times Cited Count：29 Percentile：82.31(Optics)The 
and 
3p-4d inner shell excitation lines in addition to 
2p-3s lines have been identified from the spectrum taken by an upgraded high-resolution X-ray spectrometer. It is found from analysis of the absolute intensities of the and lines that W and Mo concentrations are in the range of 
and 
, respectively, with a ratio of 
5% for ELMy H-mode plasmas with a plasma current of 2.0- 2.5 MA, a toroidal magnetic field of 2.7 T and a neutral beam injection power of 14-18 MW. For the purpose of checking self-consistency, it is confirmed that the W concentration determined from the line is in agreement with that from the line within 20% and that the plasma effective charge determined from the continuum of the first order reflection spectrum is also in agreement with that from the second order within 50%. Further, the determined plasma effective charge is in agreement with that determined from a visible spectroscopy, confirming that the sensitivity of the X-ray spectrometer is valid and that probably the W and the Mo concentrations are also valid.
Cooper, W. A.*; Brunetti, D.*; Faustin, J. M.*; Graves, J. P.*; Pfefferl
, D.*; Raghunathan, M.*; Sauter, O.*; Tran, T. M.*; Chapman, I. T.*; Ham, C. J.*; et al.
Nuclear Fusion, 55(6), p.063032_1 - 063032_8, 2015/05
Times Cited Count：2 Percentile：9.03(Physics, Fluids & Plasmas)An approximate model for a single fluid 3D MHD equilibrium with pure isothermal toroidal flow with imposed nested magnetic flux surfaces is proposed. It recovers the rigorous toroidal rotation equilibrium description in the axisymmetric limit. The approximation is valid under conditions of nearly rigid or vanishing toroidal rotation in regions with 3D deformation of the equilibrium flux surfaces. Bifurcated helical core equilibrium simulations of long-lived modes in the MAST device demonstrate that the magnetic structure is only weakly affected by the flow but that the 3D pressure distortion is important. The pressure is displaced away from the major axis and therefore is not as noticeably helically deformed as the toroidal magnetic flux under the subsonic flow conditions. Fast particle confinement is investigated with the VENUS code. In the presence of toroidal flow, the drift orbit equations depend on the electrostatic potential associated with the rotation and quasineutrality at lowest order in Larmor radius. When the equilibrium has 3D deformations, geometrical terms appear from the evaluation of Ohm's Law that considerably complicates the description of fast particle confinement.
Giroud, C.*; Jachmich, S.*; Jacquet, P.*; J
rvinen, A.*; Lerche, E.*; Rimini, F.*; Aho-Mantila, L.*; Aiba, Nobuyuki; Balboa, I.*; Belo, P.*; et al.
Plasma Physics and Controlled Fusion, 57(3), p.035004_1 - 035004_20, 2015/03
Times Cited Count：64 Percentile：95.98(Physics, Fluids & Plasmas)This paper reports the progress made at JET-ILW on integrating the requirements of the reference ITER baseline scenario with normalized confinement factor of 1, at a normalized pressure of 1.8 together with partially detached divertor whilst maintaining these conditions over many energy confinement times. The 2.5 MA high triangularity ELMy H-modes are studied with two different divertor configurations with D-gas injection and nitrogen seeding. The power load reduction with N seeding is reported. The relationship between an increase in energy confinement and pedestal pressure with triangularity is investigated. The operational space of both plasma configurations is studied together with the ELM energy losses and stability of the pedestal of unseeded and seeded plasmas.
Garcia, J.*; Hayashi, Nobuhiko; Baiocchi, B.*; Giruzzi, G.*; Honda, Mitsuru; Ide, Shunsuke; Maget, P.*; Narita, Emi*; Schneider, M.*; Urano, Hajime; et al.
Nuclear Fusion, 54(9), p.093010_1 - 093010_13, 2014/09
Times Cited Count：38 Percentile：86.74(Physics, Fluids & Plasmas)Nakano, Tomohide; Shumack, A. E.*; Maggi, C.*; Reinke, M.*; Lawson, K. D.*; Ptterich, T.*; Brezinsek, S.*; Lipschultz, B.*; Matthews, G.*; Chernyshova, M.*; et al.
Europhysics Conference Abstracts (Internet), 38F, p.P1.019_1 - P1.019_4, 2014/06
no abstracts in English
Garcia, J.*; Hayashi, Nobuhiko; Giruzzi, G.*; Schneider, M.*; Joffrin, E.*; Ide, Shunsuke; Sakamoto, Yoshiteru; Suzuki, Takahiro; Urano, Hajime; JT-60 Team; et al.
Europhysics Conference Abstracts (Internet), 38F, p.P1.029_1 - P1.029_4, 2014/06
Luce, T. C.*; Challis, C. D.*; Ide, Shunsuke; Joffrin, E.*; Kamada, Yutaka; Politzer, P. A.*; Schweinzer, J.*; Sips, A. C. C.*; Stober, J.*; Giruzzi, G.*; et al.
Nuclear Fusion, 54(1), p.013015_1 - 013015_15, 2013/12
Times Cited Count：33 Percentile：83.58(Physics, Fluids & Plasmas)Giruzzi, G.*; Garcia, J.*; Hayashi, Nobuhiko; Schneider, M.*; Artaud, J. F.*; Baruzzo, M.*; Bolzonella, T.*; Farina, D.*; Figini, L.*; Fujita, Takaaki; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
Garcia, J.*; Hayashi, Nobuhiko; Giruzzi, G.*; Schneider, M.*; Joffrin, E.*; Ide, Shunsuke; Sakamoto, Yoshiteru; Suzuki, Takahiro; Urano, Hajime; JT-60 Team; et al.
Europhysics Conference Abstracts (Internet), 36F, p.P5.057_1 - P5.057_4, 2012/00
Wiesen, S.*; Brezinsek, S.*; Jrvinen, A.*; Eich, T.*; Fundamenski, W.*; Huber, A.*; Parail, V.*; Corrigan, G.*; Hayashi, Nobuhiko; JET-EFDA Contributors*
Plasma Physics and Controlled Fusion, 53(12), p.124039_1 - 124039_12, 2011/12
Times Cited Count：22 Percentile：67.34(Physics, Fluids & Plasmas)Urano, Hajime; Saibene, G.*; Oyama, Naoyuki; Parail, V.*; de Vries, P.*; Sartori, R.*; Kamada, Yutaka; Kamiya, Kensaku; Loarte, A.*; L
nnroth, J.*; et al.
Nuclear Fusion, 51(11), p.113004_1 - 113004_10, 2011/11
Times Cited Count：10 Percentile：41.01(Physics, Fluids & Plasmas)The effect of TF ripple on the edge pedestal characteristics are examined in JET and JT-60U. By the installation of ferritic inserts, TF ripple was reduced from 
to 
in JT-60U. In JET, TF ripple was varied from 
to by feeding different currents to TF coils. The pedestal pressure was similar with reduced ripple in JT-60U. In JET, no clear difference of the pedestal characteristics was also observed. The edge toroidal rotation clearly decreased in counter direction by increased TF ripple. However, in JT-60U, the ELM frequency decreased by 
and the increased ELM loss power by 
with reduced ripple. In JET, ELM frequency increases only slightly with increased TF ripple. From this inter-machine experiment, TF ripple less than does not strongly affect the pedestal pressure. The effect of TF ripple on pedestal characteristics at lower collisionality close to ITER should be investigated as a next step study.
Imbeaux, F.*; Citrin, J.*; Hobirk, J.*; Hogeweij, G. M. D.*; Kchl, F.*; Leonov, V. M.*; Miyamoto, Seiji; Nakamura, Yukiharu*; Parail, V.*; Pereverzev, G. V.*; et al.
Nuclear Fusion, 51(8), p.083026_1 - 083026_11, 2011/08
Times Cited Count：35 Percentile：80.59(Physics, Fluids & Plasmas)Litaudon, X.*; Sakamoto, Yoshiteru; de Vries, P. C.*; Salmi, A.*; Tala, T.*; Angioni, C.*; Benkadda, S.*; Beurskens, M. N. A.*; Bourdelle, C.*; Brix, M.*; et al.
Nuclear Fusion, 51(7), p.073020_1 - 073020_13, 2011/07
Times Cited Count：8 Percentile：34.5(Physics, Fluids & Plasmas)A variety of triggering mechanisms and structures of internal transport barrier (ITB) has been observed in various devices or depending on operation scenarios. Thus identity experiments on ITB in JT-60U and JET have been performed to shed light on the physics behind ITBs. Because of their similar size, the dimensionless parameters between both devices are the same. These experiments were performed with near identical magnetic configurations, heating waveforms and normalized quantities such as safety factor, magnetic shear, normalized Larmor radius, normalized collision frequency, beta, temperatures ratio. Similarities of the ITB triggering mechanism and the ITB strength have been observed when a proper match is achieved of the most relevant profiles of the normalized quantities. This paper will report on the detail comparison of transport properties of ITBs obtained in these JET/JT-60U identity experiments.
Hayashi, Nobuhiko; Parail, V.*; Koechl, F.*; Aiba, Nobuyuki; Takizuka, Tomonori; Wiesen, S.*; Lang, P.*; Oyama, Naoyuki; Ozeki, Takahisa; JET-EFDA Contributors*
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03
Imbeaux, F.*; Basiuk, V.*; Budny, R.*; Casper, T.*; Citrin, J.*; Fereira, J.*; Fukuyama, Atsushi*; Garcia, J.*; Gribov, Y. V.*; Hayashi, Nobuhiko; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03
Mayoral, M.-L.*; Bobkov, V.*; Colas, L.*; Goniche, M.*; Hosea, J.*; Kwak, J. G.*; Pinsker, R.*; Moriyama, Shinichi; Wukitch, S.*; Baity, F. W.*; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 11 Pages, 2011/03
For any given ICRF antenna design for ITER, the maximum achievable power strongly depends on the density profiles in the SOL. It has been suggested that gas injection can be used to modify the SOL profiles and thus minimize the sensitivity of the ICRF coupling to variations in the density at the edge of the confined plasma. Recently joint experiments coordinated by the ITPA were performed to characterize further this method. An increase in SOL density during gas injection led to improved coupling for all tokamaks in this multi-machine comparison. The effectiveness of using gas injection over a wide range of conditions, as a tool to tailor the edge density in front of the ICRF antennas, is documented for different gas inlet location and plasma configurations. In addition, any deleterious effects on the confinement and interaction with the antenna near-field are not investigated.
Chapman, I. T.*; Buttery, R. J.*; Coda, S.*; Gerhardt, S.*; Graves, J. P.*; Howell, D. F.*; Isayama, Akihiko; La Haye, R. J.*; Liu, Y.*; Maget, P.*; et al.
Nuclear Fusion, 50(10), p.102001_1 - 102001_7, 2010/10
Times Cited Count：52 Percentile：87.41(Physics, Fluids & Plasmas)no abstracts in English
Imbeaux, F.*; Basiuk, V.*; Budny, R.*; Casper, T.*; Citrin, J.*; Fereira, J.*; Fukuyama, Atsushi*; Garcia, J.*; Gribov, Y. V.*; Hayashi, Nobuhiko; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10
In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from several tokamaks have been analyzed by means of integrated modeling in view of determining relevant heat transport models for these operation phases. The results of these studies are presented and projections to ITER current ramp-up and ramp-down scenarios are done, focusing on the baseline inductive scenario (main heating plateau current of 15 MA). Various transport models have been tested by means of integrated modeling against experimental data from ASDEX Upgrade, C-Mod, DIII-D, JET and Tore Supra, including both Ohmic plasmas and discharges with additional heating/current drive. With using the most successful models, projections to the ITER current ramp-up and ramp-down phases are carried out. Though significant differences between models appear on the electron temperature prediction, the final q-profiles reached in the simulation are rather close.
Sips, A. C. C.*; Casper, T.*; Doyle, E. J.*; Giruzzi, G.*; Gribov, Y.*; Hobirk, J.*; Hogeweij, G. M. D.*; Horton, L. D.*; Hubbard, A. E.*; Hutchinson, I.*; et al.
Nuclear Fusion, 49(8), p.085015_1 - 085015_11, 2009/08
Times Cited Count：53 Percentile：87.31(Physics, Fluids & Plasmas)Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for 
0.23-0.33 V m
is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps 
(3) 0.85 during the ramp up to 
= 3. A rise phase with an H-mode transition is capable of achieving (3) 0.7 at the start of the FT. Operation of the H-mode reference scenario at 3 and the hybrid scenario at = 4-4.5 during the FT phase is documented, providing data for the (3) evolution after the H-mode transition and the (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept 
1.2 during the first half of the current decay, using a slow 
ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.
Kurihara, Kenichi; Lister, J. B.*; Humphreys, D. A.*; Ferron, J. R.*; Treutterer, W.*; Sartori, F.*; Felton, R.*; Brmond, S.*; Moreau, P.*; JET-EFDA Contributors*
Fusion Engineering and Design, 83(7-9), p.959 - 970, 2008/12
Times Cited Count：25 Percentile：81.47(Nuclear Science & Technology)The existing large and medium-size tokamaks are expected to explore more advanced operation scenarios toward the ITER and a future power reactor. To specify one or more solutions to keep a steady-state plasma with high performance, and to avoid plasma instabilities almost completely, a plasma control system for ITER should have two important aspects: Technical inheritance of the currently-working functions, and flexible or adaptive structure. First, we make review on the system configuration and essential functions employed in each plasma control system from the viewpoint of hardware as well as software. Second, we survey ITER control system requirements for the current CODAC design. Third, flexible structure in the plasma control system should be discussed. Finally, on the basis of the above discussion, we would like to envisage a future plasma control system for ITER and a fusion power plant.
