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Journal Articles

Inter-code comparison benchmark between DINA and TSC for ITER disruption modelling

Miyamoto, Seiji*; Isayama, Akihiko; Bandyopadhyay, I.*; Jardin, S. C.*; Khayrutdinov, R. R.*; Lukash, V.*; Kusama, Yoshinori; Sugihara, Masayoshi*

Nuclear Fusion, 54(8), p.083002_1 - 083002_19, 2014/08

 Times Cited Count:32 Percentile:82.89(Physics, Fluids & Plasmas)

Two well-established simulation codes, DINA and TSC, are compared with each other using benchmark scenarios in order to validate the ITER 2D disruption modelling by those codes. Although the simulation models employed in those two codes ought to be equivalent in the resistive time scale, it has long been unanswered whether the one of the two codes is really able to reproduce the other result correctly, since a large number of code-wise differences render the comparison task exceedingly complicated. In this paper, it is demonstrated that after simulations are set up accounting for the model differences, in general, a good agreement is attained on a notable level, corroborating the correctness of the code results. When the halo current generation and its poloidal path in the first wall are included, however, the situation is more complicated. Because of the surface averaged treatment of the magnetic field (current density) diffusion equation, DINA can only approximately handle the poloidal electric currents in the first wall that cross field lines. Validation is carried out for DINA simulations of halo current generation by comparing with TSC simulations, where the treatment of halo current dynamics is more justifiable. The particularity of each code is depicted and the consequence in ITER disruption prediction is discussed.

Journal Articles

Role of the electron temperature in the current decay during disruption in JT-60U

Shibata, Yoshihide; Isayama, Akihiko; Matsunaga, Go; Kawano, Yasunori; Miyamoto, Seiji*; Lukash, V.*; Khayrutdinov, R.*; JT-60 Team

Plasma and Fusion Research (Internet), 9(Sp.2), p.3402084_1 - 3402084_5, 2014/06

We performed the disruption simulation using DINA code to investigate the effect of the electron temperature $$T_{rm e}$$ on the plasma current decay after the initial phase of current quench (CQ). In this calculation, we used the measured $$T_{rm e}$$ profile during the initial phase of CQ. After the initial phase of CQ, we assumed that the $$T_{rm e}$$ profile does not change in time and used the value at the end of the initial phase of current quench because $$T_{rm e}$$ profile could not be measured after the initial phase of CQ. From the simulation results, it was found that the time evolution of plasma current calculated by DINA was similar to experimental one in this calculation. However, the time evolution of $$T_{rm e}$$profile in this calculation was different from the measured $$T_{rm e}$$ profile because Te after first mini-collapse rapidly decreased until the value below a measurement limit (less than 0.1 keV). Moreover, the time evolution of poloidal cross-section S calculated by DINA code was rapidly decreased although the experimental one was gradually decreased. The plasma current decay during the disruption is determined by various parameters, $$dL_{rm p}/dt$$, $$T_{rm e}$$ and S. It is necessary to evaluate the effect of $$T_{rm e}$$ profile on the plasma current decay after the initial phase of CQ by using various assumed $$T_{rm e}$$ model and DINA code.

Journal Articles

The Effect of the electron temperature and current density profiles on the plasma current decay in JT-60U disruptions

Shibata, Yoshihide; Isayama, Akihiko; Miyamoto, Seiji*; Kawakami, Sho*; Watanabe, Kiyomasa*; Matsunaga, Go; Kawano, Yasunori; Lukash, V.*; Khayrutdinov, R.*; JT-60 Team

Plasma Physics and Controlled Fusion, 56(4), p.045008_1 - 045008_8, 2014/04

 Times Cited Count:2 Percentile:15.56(Physics, Fluids & Plasmas)

In JT-60U disruption, the plasma current decay during the initial phase of current quench has been calculated by a disruption simulation code (DINA) using the measured electron temperature $$T_{rm e}$$ profile. In the case of fast plasma current decay, $$T_{rm e}$$ has a peaked profile just after thermal quench and the $$T_{rm e}$$ profile doesn't change significantly during the initial phase of current quench. On the other hand, in the case of the slow plasma current decay, the $$T_{rm e}$$ profile is border just after the thermal quench, and the $$T_{rm e}$$ profile shrinks. The results of DINA simulation show that plasma internal inductance $$L_{rm i}$$ increases during the initial phase of current quench, while plasma external inductance $$L_{rm e}$$ does not change in time. The increase of $$L_{rm i}$$ is caused by current diffusion toward the core plasma due to the decrease of $$T_{rm e}$$ in intermediate and edge regions. It is suggested that an additional heating in the plasma periphery region has the effect of slowing down plasma current decay.

Journal Articles

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

Miyamoto, Seiji; Sugihara, Masayoshi*; Shinya, Kichiro*; Nakamura, Yukiharu*; Toshimitsu, Shinichi*; Lukash, V. E.*; Khayrutdinov, R. R.*; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji*

Fusion Engineering and Design, 87(11), p.1816 - 1827, 2012/11

 Times Cited Count:14 Percentile:71.39(Nuclear Science & Technology)

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.

JAEA Reports

Studies on representative disruption scenarios, associated electromagnetic and heat loads and operation window in ITER

Fujieda, Hirobumi; Sugihara, Masayoshi*; Shimada, Michiya; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji; Neyatani, Yuzuru

JAEA-Research 2007-052, 115 Pages, 2007/07

JAEA-Research-2007-052.pdf:3.58MB

Impacts of plasma disruptions on ITER have been investigated to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load. Heat load on the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code. For vertical displacement event, beryllium ($$Be$$) wall will not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper $$Be$$ wall and tungsten baffle due to the TQ after the vertical movement. However, its impact could be mitigated by implementing a reliable detection system of the vertical movement and a mitigation system.

Journal Articles

Progress in the ITER physics basis, 3; MHD stability, operational limits and disruptions

Hender, T. C.*; Wesley, J. C.*; Bialek, J.*; Bondeson, A.*; Boozer, A. H.*; Buttery, R. J.*; Garofalo, A.*; Goodman, T. P.*; Granetz, R. S.*; Gribov, Y.*; et al.

Nuclear Fusion, 47(6), p.S128 - S202, 2007/06

 Times Cited Count:879 Percentile:100(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Disruption scenarios, their mitigation and operation window in ITER

Shimada, Michiya; Sugihara, Masayoshi; Fujieda, Hirobumi*; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

Several representative disruption scenarios are specified and disruption simulations are performed with the DINA code and EM load analyses with the 3D FEM code for these scenarios based on newly derived physics guidelines. Although some margin is confirmed in the EM loads due to induced eddy and halo currents on the in-vessel components for all of the representative scenarios, but the margin is not large. The heat load on various parts of the first wall due to vertical movements and thermal quenches is calculated. The beryllium wall will not melt during vertical movement. Melting is anticipated at the thermal quench during a VDE, though its impact could be reduced substantially by implementing a reliable detection and mitigation system, e.g., massive gas injection. With unmitigated disruptions, the loss of beryllium layer is expected to be within 30 $$mu$$m/event out of 10 mm thick beryllium first wall.

Journal Articles

Analysis of the direction of plasma vertical movement during major disruptions in ITER

Lukash, V.*; Sugihara, Masayoshi; Gribov, Y.*; Fujieda, Hirobumi*

Plasma Physics and Controlled Fusion, 47(12), p.2077 - 2086, 2005/12

 Times Cited Count:9 Percentile:31.34(Physics, Fluids & Plasmas)

Vertical directions of plasma movement after the thermal quench (TQ) of major disruptions in ITER are investigated using the predictive mode of the DINA code. Three dominant parameters in determining the direction of plasma movement are identified; (1) the rate of plasma current quench, (2) change of the internal plasma inductance li associated with the TQ and (3) the initial vertical position of plasma column before the TQ. It is shown that the reference ITER plasma moves upward after the TQ, if the current quench rate is higher than 200kA/ms and the drop of li does not exceed 0.2 for the present reference initial vertical position (55.5 cm above the center of machine).

Journal Articles

Modeling of plasma current decay during the disruption

Owaki, Hirokazu; Sugihara, Masayoshi; Kawano, Yasunori; Lukash, V. V.*; Khayrutdinov, R. R.*; Zhogolev, V.*; Ozeki, Takahisa; Hatayama, Akiyoshi*

Europhysics Conference Abstracts (CD-ROM), 29C, 4 Pages, 2005/00

no abstracts in English

Journal Articles

Edge safety factor at the onset of plasma disruption during VDEs in JT-60U

Sugihara, Masayoshi; Lukash, V.*; Khayrutdinov, R.*; Neyatani, Yuzuru

Plasma Physics and Controlled Fusion, 46(10), p.1581 - 1589, 2004/10

 Times Cited Count:11 Percentile:34.79(Physics, Fluids & Plasmas)

Detailed examinations of the value of the edge safety factor q at the onset of thermal quench during intentional VDE experiments in JT-60U are performed using two different reconstruction methods, FBI/FBEQU and DINA. Results by both methods are very similar and show that the thermal quench occurs when the q value is in the range between 1.5-2. This result suggests that the predictive simulations for VDEs should be performed within this range of q to examine the subsequent differences in the halo currents, plasma movement and other plasma behavior during the current quench.

Journal Articles

Examinations on plasma behaviour during disruptions on existing tokamaks and extrapolation to ITER

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Proceedings of 30th EPS Conference on Controlled Fusion and Plasma Physics (CD-ROM), 4 Pages, 2003/07

We examine plasma behaviours during plasma disruptions in detail in JT-60U and other tokamaks to derive appropriate physics guidelines for the behaviours. Their interpretations and their extrapolations to ITER are incorporated into the DINA code, which solves plasma transport and 2D free boundary plasma equilibrium simultaneously with circuit equations for the vacuum vessel and the PF coils. Sensitivity of the plasma behaviours and their impact on the EM force during disruptions due to the range of variation and uncertainty of the experimental data are examined.

Journal Articles

Wave form of current quench during disruptions in Tokamaks

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Purazuma, Kaku Yugo Gakkai-Shi, 79(7), p.706 - 712, 2003/07

The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential.

Journal Articles

Runaway current termination in JT-60U

Tamai, Hiroshi; Yoshino, Ryuji; Tokuda, Shinji; Kurita, Genichi; Neyatani, Yuzuru; Bakhtiari, M.; Khayratdinov, R. R.*; Lukash, V.*; Rosenbluth, M. N.*; JT-60 Team

Nuclear Fusion, 42(3), p.290 - 294, 2002/03

 Times Cited Count:34 Percentile:70.79(Physics, Fluids & Plasmas)

no abstracts in English

Oral presentation

Analysis of current quench of the JT-60U tokamak by using two dimensional MHD equilibrium calculation code (DINA)

Kawakami, Sho*; Shibata, Yoshihide*; Watanabe, Kiyomasa*; Ono, Noriyasu*; Kajita, Shin*; Okamoto, Masaaki*; Isayama, Akihiko; Sugihara, Masayoshi*; Kawano, Yasunori; Lukash, V. E.*; et al.

no journal, , 

no abstracts in English

Oral presentation

Analysis of neon gas-puff disruption in the JT-60U tokamak by using DINA disruption simulation code

Kawakami, Sho*; Ono, Noriyasu*; Watanabe, Kiyomasa*; Shibata, Yoshihide; Okamoto, Masaaki*; Miyamoto, Seiji; Isayama, Akihiko; Sugihara, Masayoshi*; Kawano, Yasunori; Lukash, V. E.*; et al.

no journal, , 

no abstracts in English

Oral presentation

Analysis of current quench induced by the massive neon gas-puff in JT-60U using DINA code

Shibata, Yoshihide; Isayama, Akihiko; Matsunaga, Go; Kawano, Yasunori; Miyamoto, Seiji*; Lukash, V. E.*; Khayrutdinov, R.*

no journal, , 

no abstracts in English

19 (Records 1-19 displayed on this page)
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