Refine your search:     
Report No.
 - 
Search Results: Records 1-6 displayed on this page of 6
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Numerical study of plasma generation process and internal antenna heat loadings in J-PARC RF negative ion source

Shibata, Takanori*; Nishida, Kenjiro*; Mochizuki, Shintaro*; Mattei, S.*; Lettry, J.*; Hatayama, Akiyoshi*; Ueno, Akira; Oguri, Hidetomo; Okoshi, Kiyonori; Ikegami, Kiyoshi*; et al.

Review of Scientific Instruments, 87(2), p.02B128_1 - 02B128_3, 2016/02

BB2015-1473.pdf:4.28MB

 Times Cited Count:3 Percentile:16.61(Instruments & Instrumentation)

A numerical model of plasma transport and electromagnetic field in the J-PARC RF ion source has been developed to understand relation between antenna coil heat loadings and plasma production/transport processes. From the calculation, the local plasma density increase is observed in the region close to the antenna coil. The magnetic field line with absolute magnetic flux density 30-120 Gauss results in the magnetization of electron which leads to high local ionization rate. The results suggest that modification of magnetic configuration can be made to reduce plasma heat flux onto the antenna.

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

Imbeaux, F.*; Citrin, J.*; Hobirk, J.*; Hogeweij, G. M. D.*; K$"o$chl, 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:80.68(Physics, Fluids & Plasmas)

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

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

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

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.

Journal Articles

Experimental studies of ITER demonstration discharges

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:52 Percentile:87.07(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 $$E$$$$_{axis}$$ $$<$$ 0.23-0.33 V m$$^{-1}$$ 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 $$l$$$$_{i}$$(3) $$<$$ 0.85 during the ramp up to $$q$$$$_{95}$$ = 3. A rise phase with an H-mode transition is capable of achieving $$l$$$$_{i}$$(3) $$<$$ 0.7 at the start of the FT. Operation of the H-mode reference scenario at $$q$$$$_{95}$$ $$sim$$ 3 and the hybrid scenario at $$q$$$$_{95}$$ = 4-4.5 during the FT phase is documented, providing data for the $$l$$$$_{i}$$(3) evolution after the H-mode transition and the $$l$$$$_{i}$$(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 $$leq$$ 1.2 during the first half of the current decay, using a slow $$I$$$$_{p}$$ 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.

Journal Articles

Experimental studies of ITER demonstration discharges

Sips, A. C. C.*; Casper, T. A.*; Doyle, E. J.*; Giruzzi, G.*; Gribov, Y.*; Hobirk, J.*; Hogeweij, G. M. D.*; Horton, L. D.*; Hubbard, A. E.*; Hutchinson, I.*; et al.

Proceedings of 22nd IAEA Fusion Energy Conference (FEC 2008) (CD-ROM), 8 Pages, 2008/10

The ITER discharge evolution has been verified in dedicated experiments. Results show that breakdown at E$$<$$ 0.23-0.32 V/m is possible un-assisted (ohmic) for large devices like JET and attainable in all devices with 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. Operation of the H-mode reference scenario at q$$_{95}$$ = 3 and the hybrid scenario at q95=4-4.5 during the flat top phase was documented. Specific studies during the flat top phase provide data for the li evolution after the H-mode transition and the li evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation.

6 (Records 1-6 displayed on this page)
  • 1