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Journal Articles

Dust behavior during laser cutting in huge water tank

Miyamoto, Yuta; Uemura, Masaru*; Yoshikawa, Katsuhiro*; Ando, Seiji*

Reiwa-2-Nendo Koeki Zaidan Hojin Wakasawan Enerugi Kenkyu Senta kenkyu Nempo, 23, P. 40, 2021/10

no abstracts in English

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:26 Percentile:83.69(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:12.49(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:11 Percentile:69.17(Nuclear Science & Technology)

Journal Articles

Effects of ELM mitigation coils on energetic particle confinement in ITER steady-state operation

Tani, Keiji*; Shinohara, Koji; Oikawa, Toshihiro*; Tsutsui, Hiroaki*; Miyamoto, Seiji; Kusama, Yoshinori; Sugie, Tatsuo

Nuclear Fusion, 52(1), p.013012_1 - 013012_21, 2012/01

 Times Cited Count:30 Percentile:82.76(Physics, Fluids & Plasmas)

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:84.28(Physics, Fluids & Plasmas)

Journal Articles

Linear response model of the vertical electromagnetic force on a vessel applicable to ITER and future tokamaks

Miyamoto, Seiji

Plasma Physics and Controlled Fusion, 53(8), p.082001_1 - 082001_7, 2011/08

 Times Cited Count:4 Percentile:21.51(Physics, Fluids & Plasmas)

Vertical displacement events (VDEs) and a subsequent plasma disruption cause severe electromagnetic force on the vacuum vessel of axisymmetric magnetic confinement fusion devices like tokamaks and spherical tokamaks. This force is a dominant factor for the supporting system of the vacuum vessel and in-vessel components and a lot of efforts have been devoted to predict the possible force in future machines such as ITER. The eddy and halo currents induced in the vessel accompanying VDE and current quench complicate the analysis of the electromagnetic force and usually, computer simulations are required to employ for the analysis. So far, a database of vertical force for ITER has been created based on DINA simulation. However, no theory has been developed for systematic explanation of the simulation data. The problem of calculating vertical electromagnetic force on the vessel is reformulated to a linear response problem. First, it is shown that a burdensome task of calculating in-vessel halo and eddy currents is reduced to the calculation of the source term of the vertical force, or a force exerted on the plasma by poloidal field (PF) coils. The calculation is carried out without relying on the knowledge of currents in the vessel. The vertical force then emerges as a result of linear response, or electromagnetic shielding by the vessel. The model provides an analytical way of calculating vertical force. As an example of application, dependence of vertical force on current quench rate is derived analytically. The obtained formula well reproduces the simulation result of DINA.

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

TSC modelling of major disruption and VDE events in NSTX and ASDEX-upgrade and predictions for ITER

Bandyopadhyay, I.*; Gerhardt, S.*; Jardin, S.*; Sayer, R. O.*; Nakamura, Yukiharu*; Miyamoto, Seiji; Pautasso, G.*; Sugihara, Masayoshi*; ASDEX Upgrade Team*; NSTX Team*

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

Vertical Displacement Events (VDEs) and Major Disruptions (MDs) of the plasma current will induce large electromagnetic forces on the ITER machine. Estimation of these forces based on accurate modeling of these events is necessary for a robust ITER design. Originally the estimates for electromagnetic forces on ITER were carried out with the help of DINA simulations. However, since simulations of these events may be significantly influenced by model assumptions of a given code it is important to validate the results against other codes like TSC, as also benchmark and update the codes with experimental data. In this paper, we present TSC modeling of the VDE and MD events in NSTX and ASDEX-U devices, which help in improving and validating the models used in the code. The predictive modeling results for ITER with the updated code, including the force predictions, are also presented.

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

TSC modelling approach to mimicking the halo current in ASDEX upgrade disruptive discharges

Nakamura, Yukiharu*; Pautasso, G.*; Sugihara, Masayoshi*; Miyamoto, Seiji; Toshimitsu, Shinichi; Yoshino, Ryuji; ASDEX Upgrade Team*

Proceedings of 37th European Physical Society Conference on Plasma Physics (EPS 2010) (CD-ROM), 4 Pages, 2010/06

Of particular importance for the assessment of electromagnetic loads on vacuum vessel and in-vessel components of ITER is the halo current which achieves a maximum during VDEs (VDE: vertical displacement event). However, halo current models have a limited development so far with a few exceptions such as a validation study of the JT-60U halo current modelling using the DINA code. Recently, several experimental groups have prepared systematic halo current data, and further model development and validation with these data need to be performed using an axisymmetric, two-dimensional, free boundary code, TSC. To enhance an understanding of the maximum halo current and large vertical shifts, a reference discharge was selected from those included in the ASDEX upgrade disruption database. Systematic TSC simulations were performed to mimic the observation of a slow VDE of hot plasma and an ensuing fast downward-going VDE during a subsequent plasma current quench. Careful parameter adjustment of the temperature and width of the halo region was examined to mimic measurements of the halo current. A spontaneous, downward-going VDE was reproduced accurately in a manner that closely resembled experimental observations.

Journal Articles

Modeling of L-H/H-L transition in TSC simulation using JT-60U experimental data

Miyamoto, Seiji; Nakamura, Yukiharu*; Hayashi, Nobuhiko; Oyama, Naoyuki; Takenaga, Hidenobu; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji

Proceedings of 36th European Physical Society Conference on Plasma Physics (CD-ROM), 4 Pages, 2009/07

The neutral dynamics including fueling, divertor pumping, charge exchange penetration, wall retention and so on would complicate the analysis of ITER plasma behavior such as H-L back transition during plasma current ramp-down. Recently, a relatively simple model of neutral dynamics was developed by us with TSC code to describe the plasma behavior during L-H and H-L transition phase. This model is compared with a JT-60U shot, in which it is possible to extract the effect of particle confinement change on neutral because H-mode discharge is switched on/off according to EC injection and thereby particle source density is kept constant during transition. It is shown that TSC simulation can account the behavior of neutral inferred from the experimental D$$_alpha$$ signal. It is concluded that this model is applicable to scenario development of the ITER.

Journal Articles

TSC simulation of ITER plasma termination scenario with stable H-L mode transition and avoidance of radiation collapse

Nakamura, Yukiharu*; Miyamoto, Seiji; Toshimitsu, Shinichi; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji

Proceedings of 36th European Physical Society Conference on Plasma Physics (CD-ROM), 4 Pages, 2009/07

The ITER termination scenario from 15 MA to 1.5 MA (500 s $$<$$ t $$leq$$ 700 s) was reviewed by self-consistent simulations with the TSC code, comprised of newly developed D-T fuelling and pumping-out system. At 600 s, when the plasma current decreased to 10 MA, auxiliary NB heating was switched off to cease fusion $$alpha$$-heating. Simultaneously, the energy confinement switches H to L mode by intentionally removing the H mode pedestal of edge transport barrier. The H to L mode transition dynamics, ${it e.g.}$ reduction in the plasma density while building-up of in-vessel neutral gas, disappearance of the edge BS current and consequent jump in the internal inductance $$l_i(3)$$, were investigated to assess performance of the ITER pump-out system. It was newly shown that the forced H to L mode transition may trigger a radiation collapse, consequently terminating the discharge. It was also demonstrated that EC heating with 170 GHz O-mode wave after the H to L mode transition provides an effective control means to hedge risk of the radiation collapse.

Journal Articles

Thickness distribution of high-speed free-surface lithium flow simulating IFMIF target

Kondo, Hiroo*; Kanemura, Takuji*; Sugiura, Hirokazu*; Yamaoka, Nobuo*; Miyamoto, Seiji*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; Horiike, Hiroshi*

Fusion Engineering and Design, 84(7-11), p.1086 - 1090, 2009/06

 Times Cited Count:7 Percentile:48.67(Nuclear Science & Technology)

A liquid lithium(Li) target of International Fusion Materials Irradiation Facility (IFMIF) is formed as flat plane free-surface flow by a nozzle and flows at high speed around 15 m/s. This paper focuses on flatness of the liquid Li target. A Li flow experiment was conducted in Osaka University Li Loop with a test section which was 1/2.5 scaled model of IFMIF. A thickness of the Li flow was measured and obtained by a contact method which was developed for the measurement. Analytical study on Kelvin wake and numerical calculation on wakes near side walls of the flow channel were also conducted and compared with the experimental results. As the results, positions of wake crest obtained from both of the experiment and numerical calculation assuming contact angle 140$$^{circ}$$ agreed well with an iso-phase line of the analytical model. Generation of the wake are likely depends on wettability between Li and a structural material which is 304SS in the present study.

Journal Articles

Development of velocity measurement on a liquid lithium flow for IFMIF

Sugiura, Hirokazu*; Kondo, Hiroo*; Kanemura, Takuji*; Niwa, Yuta*; Yamaoka, Nobuo*; Miyamoto, Seiji*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; et al.

Fusion Engineering and Design, 84(7-11), p.1803 - 1807, 2009/06

 Times Cited Count:3 Percentile:26.15(Nuclear Science & Technology)

To develop a diagnostics system in view of its application on International Fusion Materials Irradiation Facility (IFMIF) liquid lithium (Li) target, velocity measurements on a liquid Li flow were performed in a Li circulation loop at Osaka University with a test section having a contraction nozzle 1/2.5 scale of IFMIF and producing a flat plane jet of 70 mm width and 10 mm thickness. Based on the Particle Image Velocimetry (PIV) technique, a local Li flow velocity distribution was measured by tracking brightness intensity patterns of surface waves generated on the flow. Measured surface velocity showed good agreement with a surface velocity obtained in previous water experiments, and had an insignificant effect at an area corresponding to a deuteron beam irradiation area on the IFMIF target.

Journal Articles

Measurement of free surface of liquid metal lithium jet for IFMIF target

Kondo, Hiroo*; Kanemura, Takuji*; Yamaoka, Nobuo*; Miyamoto, Seiji*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; Horiike, Hiroshi*

Fusion Engineering and Design, 82(15-24), p.2483 - 2489, 2007/10

 Times Cited Count:7 Percentile:49.85(Nuclear Science & Technology)

Lithium flow experiments were conducted for International Fusion Materials Irradiation Facility (IFMIF) at Osaka University. In the experiment, Li plane jet of 10 mm in depth and 70 mm in width formed by a two contractions nozzle was tested in the velocity range of less than 15 m/s. In the present report, Li surface measurement by pattern projection method was tested. This is a three dimensional image measurement, where stripe patterns are projected onto the flow surface without touching it. The projected patterns were observed to be deformed according to the surface up- and- down. Three-dimensional surface shape could be obtained by analyzing the deformed patterns. By the method, shapes of wave pattern called surface wakes were successfully measured. The surface wakes were observed to be formed from the nozzle edge. It was found that the nozzle edge was damaged and became serrated after lithium flowing of 1,300 hours at this moment.

Journal Articles

Investigation of free-surface fluctuations of liquid lithium flow for IFMIF lithium target by using an electro-contact probe

Kanemura, Takuji*; Kondo, Hiroo*; Yamaoka, Nobuo*; Miyamoto, Seiji*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; Horiike, Hiroshi*

Fusion Engineering and Design, 82(15-24), p.2550 - 2557, 2007/10

 Times Cited Count:21 Percentile:81.84(Nuclear Science & Technology)

For a study on characteristics of lithium target flow of International Fusion Materials Irradiation Facility (IFMIF), experiments were carried out by using a lithium loop at Osaka University. In the experiment, fluctuations of a free surface of the horizontal flow were directly measured by using an electro-contact probe acquiring condition of contact/non-contact of the probe with the flow surface as voltage data. Vertical location of the probe tip was set by 0.1 mm step. Horizontal location of the probe was 175 mm downstream from the nozzle exit, corresponding to the footprint of deuteron beam in the IFMIF case. It was found that the maximum amplitude of the surface wave, including rarely arising ones, was 2.2 mm at the center of the flow channel with width of 70 mm at the maximum flow velocity of 15 m/s. The average thickness of the flow was found to be 10.13 mm.

Journal Articles

Free-surface fluctuation at high speed lithium flow for IFMIF

Horiike, Hiroshi*; Kondo, Hiroo*; Nakamura, Hiroo; Miyamoto, Seiji*; Yamaoka, Nobuo*; Matsushita, Izuru*; Ida, Mizuho; Ara, Kuniaki; Muroga, Takeo*; Matsui, Hideki*

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

no abstracts in English

Journal Articles

Surface wave on high speed liquid lithium flow for IFMIF

Kondo, Hiroo*; Fujisato, Atsushi*; Yamaoka, Nobuo*; Inoue, Shoji*; Miyamoto, Seiji*; Iida, Toshiyuki*; Nakamura, Hiroo; Ida, Mizuho*; Matsushita, Izuru*; Muroga, Takeo*; et al.

Fusion Engineering and Design, 75-79, p.865 - 869, 2005/11

 Times Cited Count:20 Percentile:79.69(Nuclear Science & Technology)

no abstracts in English

53 (Records 1-20 displayed on this page)