Shimada, Michiya; Miyazawa, Junichi*
Purazuma, Kaku Yugo Gakkai-Shi, 92(2), p.119 - 124, 2016/02
Actively convected liquid metal divertor is promising for providing a solution for issues of DEMO reactors including heat removal and disruptions. This chapter gives an overview of the motivation, research history, recent development, future perspective and issues to be resolved.
McCracken, G.*; Stott, P.*; Iiyoshi, Atsuo*; Muraoka, Katsunori*; Nakai, Sadao*; Shimada, Michiya
Kaku Yugo; Uchu No Enerugi O Watashitachi No Te Ni, 326 Pages, 2015/00
Fusion; The Energy of the Universe, 2e is an essential reference providing basic principles of fusion energy from its history to the issues and realities progressing from the present day energy crisis. The book provides detailed developments and applications for researchers entering the field of fusion energy research. This second edition includes the latest results from the National Ignition Facility at the Lawrence Radiation Laboratory at Livermore, CA, and the progress on the International Thermonuclear Experimental Reactor (ITER) tokamak programme at Caderache, France.
Shimada, Michiya; Hirooka, Yoshihiko*
Nuclear Fusion, 54(12), p.122002_1 - 122002_7, 2014/12
Tungsten is considered to be the most promising material for divertor in a fusion reactor. Tungsten divertor can withstand the heat loads of ITER, but the heat loads of DEMO divertor is a challenge. Pulsive heat loads as those associated with disruption could melt tungsten targets. The surface would not be flat after subsequent resolidification, which would significantly deteriorate its heat handling capability. Furthermore, DBTT of tungsten is rather high: 400C, which would become even higher after neutron irradiation, possibly resulting in cracks in tungsten. Our proposal is to use liquid metal for the divertor target material and actively circulate it with force. A simplified analysis of mhd equation in a cylindrical geometry suggests that the engineering requirement is modest. This analysis suggests that this new divertor concept merits further investigation.
Shimada, Michiya; Hirooka, Yoshihiko*; Zhou, H.*
Europhysics Conference Abstracts (Internet), 38F, p.O2.110_1 - O2.110_4, 2014/00
Tungsten is considered to be most promising candidate for divertor target material for fusion reactor. Although tungsten target can withstand the heat loads of ITER, the heat exhaust requirement for DEMO is much more demanding. Pulsive heat loads associated with disurption would melt the tungsten divertor target. Melting and subsequent resolidification will roughen the tungsten surface, significantly deteriorating the heat handling capability. Further, tungsten has a rather high DBTT (Ductile-Brittle-Transition temperature) of 400C. Neutron irradiation would further increase the DBTT, which could result in cracks. In view of these issues, liquid metal divertor is proposed, which is actively circulated with the Lorentz force introduced through the electrodes in the liquid metal. A modest flow speed of 0.3 m/s seems to be adequate for the heat load exhaust of DEMO. A simple treatment of MHD equation in a cylindrical geometry suggests that the requirements on the current and voltage are modest if the ramp-up of current is made slowly (e.g. in a minute), implying that the this concept is worth further study.
Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.
Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12
This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.
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
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 () wall will not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper 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.
Shimada, Michiya; Campbell, D. J.*; Mukhovatov, V.*; Fujiwara, Masami*; Kirneva, N.*; Lackner, K.*; Nagami, Masayuki; Pustovitov, V. D.*; Uckan, N.*; Wesley, J.*; et al.
Nuclear Fusion, 47(6), p.S1 - S17, 2007/06
The Progress in the ITER Physics Basis document is an update of the ITER Physics Basis (IPB), which was published in 1999. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modeling and theoretical activities carried out on today's tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the International Tokamak Physics Activities (ITPA), which were initiated by the European Union, Japan, Russia and the U.S.A.. The new methodologies of projection and control developed through the ITPA are applied to ITER, which was redesigned under revised technical objectives, but will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy.
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
no abstracts in English
Gribov, Y.*; Humphreys, D. A.*; Kajiwara, Ken*; Lazarus, E. A.*; Lister, J. B.*; Ozeki, Takahisa; Portone, A.*; Shimada, Michiya*; Sips, A. C. C.*; Wesley, J. C.*
Nuclear Fusion, 47(6), p.S385 - S403, 2007/06
This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This chapter considers only plasma initiation and plasma basic control. The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, ITER can produce plasma initiation in a low toroidal electric field of 0.3V/m, if it is assisted by about 2MW of ECRF heating. The plasma basic control is described, which includes control of the plasma current, position and shape - the plasma magnetic control, as well as control of other plasma global parameters or their profiles - the plasma performance control.
Ioki, Kimihiro; Chuyanov, V.*; Elio, F.*; Garkusha, D.*; Gribov, Y.*; Lamzin, E.*; Morimoto, Masaaki; Shimada, Michiya; Sugihara, Masayoshi; Terasawa, Atsumi; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
Two important design updates have been made in the ITER VV and in-vessel components recently. One is the introduction of limiters moveable during a plasma discharge, and the other is optimization of the ferromagnetic insert configuration to minimize the toroidal field ripple. In the new limiter concept, the limiters are retracted by 8 cm during the plasma flat top phase in the divertor configuration. This concept gives important advantages: (1) the particle and heat loads due to disruptions, ELMs and blobs on the limiters will be mitigated approximately by a factor 1.5 or more; (2) the gap between the plasma and the ICRH antenna can be reduced to improve the coupling of the ICRH power. The ferromagnetic inserts have previously not been planned to be installed in the outboard midplane region between equatorial ports due to irregularity of tangential ports for NB injection. The result is a relatively large ripple (1 %) in a limited region of the plasma, which nevertheless seems acceptable from the plasma performance viewpoint. However, toroidal field flux lines fluctuate 10 mm due to the large ripple in the FW region. To avoid problems due to the large TF flux line fluctuation, additional ferromagnetic inserts are now planned to be installed in the equatorial port region.
Oikawa, Toshihiro; Shimada, Michiya; Polevoi, A. R.*; Naito, Osamu; Bonoli, P. T.*; Hayashi, Nobuhiko; Kessel, C. E.*; Ozeki, Takahisa
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 7 Pages, 2007/03
The current drive capability of lower hybrid waves is assessed for an ITER steady state scenario using a relativistic, one-dimensional Fokker-Planck code and a toroidal ray tracing code. The present LH launcher design provides a current drive efficiency of 1.810AWm and an off-axis profile of the driven current that is fovarable for a reversed magnetic shear configuration. Possible optimizations in the LH power spectrum are investigated. Neutral beam current drive (NBCD) is investigated with theoretical codes employing different approaches. An NBCD code employing the bounce-averaged Fokker-Planck equation to include orbit effects in a toroidal system shows a good agreement with an orbit following Monte-Carlo code.
Kamada, Yutaka; Leonard, A. W.*; Bateman, G.*; Becoulet, M.*; Chang, C. S.*; Eich, T.*; Evans, T. E.*; Groebner, R. J.*; Guzdar, P. N.*; Horton, L. D.*; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
no abstracts in English
Polevoi, A. R.*; Shimada, Michiya; Sugihara, Masayoshi; Igitkhanov, Y. L.*; Mukhovatov, V.*; Kukushkin, A. S.*; Medvedev, S. Y.*; Zvonkov, A. V.*; Ivanov, A. A.*
Nuclear Fusion, 45(11), p.1451 - 1456, 2005/11
Requirements for pellet injection parameters for plasma fuelling are assessed for ITER scenarios with enhanced particle confinement. A pellet injection throughput of 100 Pam/s would be sufficient. The assessment is based on the integrated transport simulations including models of pedestal transport, reduction of helium transport and boundary conditions compatible with SOL/divertor simulations. The requirements for pellet injection for the inductive H-mode scenario (HH98(y,2) = 1) are reconsidered taking account of a possible reduction of the particle loss obtained in some experiments at low collisionalities. The assessment of fuelling requirements is carried out for the hybrid and steady state scenarios with enhanced confinement with HH98(y,2) 1. A robustness of plasma performance to the variation of particle transport is demonstrated. A new type of steady state (SS) scenario is considered with neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD) instead of lower hybrid current drive (LHCD).
Kamada, Yutaka; Shimada, Michiya; Miura, Yukitoshi; Ogawa, Yuichi*
Purazuma, Kaku Yugo Gakkai-Shi, 81(11), p.849 - 862, 2005/11
This section discusses fusion plasma research that needs to be carried out to develop fusion power plants. Burning plasma, in which self-heating by energetic alpha particles plays an essential role, should be recognized as autonomous system. This is quite different from present plasma experiments, suggesting a possibility to yield some qualitative changes in fusion plasma research. Research with ITER is strongly expected to contribute to this burning plasma physics. In addition, plasma performance in steady-state and at high beta is very important in fusion power plants from the engineering and economical viewpoints. Plasma parameters expected for fusion power plants are discussed, and present status of experimental research is reviewed. Research in devices other than ITER with unique features would be instrumental for exploring high performance plasmas. A necessity of research complementary to ITER plasma is discussed.
Shimada, Michiya; Costley, A. E.*; Federici, G.*; Ioki, Kimihiro*; Kukushkin, A. S.*; Mukhovatov, V.*; Polevoi, A. R.*; Sugihara, Masayoshi
Journal of Nuclear Materials, 337-339, p.808 - 815, 2005/03
ITER is an experimental fusion reactor for investigation and demonstration of burning plasmas, characterised of its heating dominated by alpha-particle heating. ITER is a major step from present devices and an indispensable step for fusion reactor development. ITER's success largely depends on the control of plasma-wall interactions(PWI), with power and particle fluxes and time scales one or two orders of magnitude larger than in present devices. The strategy for control of PWI includes the semi-closed divertor, strong fuelling and pumping, disruption and ELM control, replaceable plasma-facing materials and stepwise operation.
Shimada, Michiya; Campbell, D.*; Stambaugh, R.*; Polevoi, A. R.*; Mukhovatov, V.*; Asakura, Nobuyuki; Costley, A. E.*; Donn, A. J. H.*; Doyle, E. J.*; Federici, G.*; et al.
Proceedings of 20th IAEA Fusion Energy Conference (FEC 2004) (CD-ROM), 8 Pages, 2004/11
This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors 4. Extrapolation to ITER suggests that at the reduced plasma current of 12MA, high Q 10 and long pulse (1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces.
Asakura, Nobuyuki; Iio, Shunji*; Ozeki, Takahisa; Ono, Yasushi*; Kato, Takako*; Kawano, Yasunori; Sugihara, Masayoshi; Takamura, Shuichi*; Tanabe, Tetsuo*; Nakajima, Noriyoshi*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 80(7), P. 642, 2004/07
no abstracts in English
Purazuma, Kaku Yugo Gakkai-Shi, 80(3), p.222 - 226, 2004/03
Discussion is made on objectives of edge plasma and plasma-wall interaction studies, divertor performance projection of ITER and key R&D issues on plasma-wall interaction including material issues, tritium retention, and transient events like ELMs and disruptions. A perspective of future development is discussed for the purpose of projection to and control of ITER plasma.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Recent performance analysis has improved confidence in achieving Q 10 in inductive operation in ITER. Performance analysis based on empirical scaling shows the feasibility of achieving Q 10 in inductive operation with a sufficient margin. Theory-based core modeling indicates the need of high pedestal temperature (2-4 keV) to achieve Q 10, which is in the range of projection with pedestal scaling. The heat load of type-I ELM could be made tolerable by high density operation and further tilting the target plate (if necessary). Pellet injection from High-Field Side would be useful in enhancing Q and reducing ELM heat load. Steady state operation scenarios have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A. S.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Performance analysis based on empirical scaling shows the feasibility of achieving Q 10 in inductive operation. Analysis has also elucidated a possibility that ITER can potentially demonstrate Q 50, enabling studies of self-heated plasmas. Theory-based core modeling indicates the need of high pedestal temperature (3.2 - 5.3 keV) to achieve Q 10, which is in the range of projection with presently available pedestal scalings. Pellet injection from high-field side would be useful in enhancing Q and reducing ELM heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.