高温プラズマにおける水素原子挙動および粒子閉じ込めに関する研究

第1章 緒言 第2章 高温プラズマ中の水素原子挙動の研究手法 第3章 水素原子挙動および粒子閉じ込め研究手法の開発 第4章 高温プラズマにおける粒子閉じ込め 第5章 結言Made available in DSpace on 2012-07-04T00:33:51Z (GMT). No. of bitstreams: 1 takenaga.pdf: 14103569 bytes, checksum: 8e365a7c292cdc81c244d8636b2c011d (MD5) Previous issue date: 1995-03-2

原型炉のための技術基盤確立に向けた日本の取組

The establishment of technology bases required for the development of a fusion demonstration reactor (DEMO) has been discussed by a joint effort throughout the Japanese fusion community. The basic concept of DEMO premised for investigation has been identified and the structure of technological issues to ensure the feasibility of this DEMO concept has been examined. The Joint-Core Team, which was launched along with the request by the ministerial council, has compiled analyses in two reports to clarify technology which should be secured, maintained, and developed in Japan, to share the common targets among industry, government, and academia, and to activate actions under a framework for implementation throughout Japan. The reports have pointed out that DEMO should be aimed at steady power generation beyond several hundred thousand kilowatts, availability which must be extended to commercialization, and overall tritium breeding to fulfill self-sufficiency of fuels. The necessary technological activities, such as superconducting coils, blanket, divertor, and others, have been sorted out and arranged in the chart with the time line toward the decision on DEMO. Based upon these Joint-Core Team reports, related actions are emerging to deliberate the Japanese fusion roadmap

Development of Strategic Establishment of Technology Bases for a Fusion DEMO Reactor in Japan

The strategic establishment of technology bases required for the development of a fusion demonstration reactor (DEMO) has been discussed by joint endeavors throughout the Japanese fusion community. The mission of Fusion DEMO is to demonstrate the technological and economic feasibility of fusion energy. The basic concept of Fusion DEMO has been identified and the structure of technological issues to ensure the feasibility of this DEMO concept has been examined. The Joint-Core Team consisting of experts from the Japanese fusion community including industry has pointed out that DEMO should be aimed at steady power generation beyond several hundred thousand kilowatts, availability which must be extensible to commercialization, and overall tritium breeding sufficient to achieve fuel-cycle self-sufficiency. The necessary technological issues and activities have been sorted out along with 11 identified elements of DEMO, such as superconducting coils, blanket, divertor, and others. These will be arranged within a time line to lead to the Japanese fusion roadmap

Recent Results from LHD Experiment with Emphasis on Relation to Theory from Experimentalist’s View

he Large Helical Device (LHD) has been extending an operational regime of net-current free plasmas towardsthe fusion relevant condition with taking advantage of a net current-free heliotron concept and employing a superconducting coil system. Heating capability has exceeded 10 MW and the central ion and electron temperatureshave reached 7 and 10 keV, respectively. The maximum value of β and pulse length have been extended to 3.2% and 150 s, respectively. Many encouraging physical findings have been obtained. Topics from recent experiments, which should be emphasized from the aspect of theoretical approaches, are reviewed. Those are (1) Prominent features in the inward shifted configuration, i.e., mitigation of an ideal interchange mode in the configuration with magnetic hill, and confinement improvement due to suppression of both anomalous and neoclassical transport, (2) Demonstration ofbifurcation of radial electric field and associated formation of an internal transport barrier, and (3) Dynamics of magnetic islands and clarification of the role of separatrix

Neural-network-based semi-empirical turbulent particle transport modelling founded on gyrokinetic analyses of JT-60U plasmas

Novel turbulent particle transport modelling has been proposed following joint analyses with gyrokinetic calculations and JT-60U experimental data. Here the diagonal (diffusion) and off- diagonal (pinch) transport mechanisms are treated individually. Besides the decomposition, realistic particle sources from neutral-beam fuelling, which have not been discussed in earlier gyrokinetic studies on particle transport, are taken into account. Taking advantage of the features offered by the modelling, the contribution from each transport mechanism to the turbulent particle flux has been quantitatively clarified. Furthermore, a framework has been developed to calculate the turbulent particle flux driven by each transport mechanism accurately and quickly, taking a neural-network-based approach. The framework can be used for fast prediction of density profiles and for investigating the effects of the transport mechanisms on density profile formation

Scaling study for positive magnetic shear ELMy H-mode plasmas

Scaling of the density peaking for positive magnetic shear ELMy H-mode plasmas in JT-60U has been developed. Although the density peaking generally depends on collisionality, a variation of the density peaking factor for the same collisionality exists, and the variation is different between plasmas with co-directed and ctr-directed toroidal rotation(co-V T plasmas and ctr-V T plasmas). The variation of the co-V T plasmas can be explained by particle source profile. As a result of scaling, the peaking factor of the co-V T plasmas depends on the particle source rate profile from neutral beams (NBs) as much as collisionality. The dataset of the ctr-V T plasmas has a larger variation of the peaking factor compared to that of the co-V T plasmas. The larger variation stems from that the peaking factor of the ctr-V T plasmas also depends on the normalized ion temperature gradient at the edge region. As a result of scaling, the normalized ion temperature gradient influences the peaking factor as much as or a little larger than collisionality. The parameter regime of the ctr-V T plasmas ranges from the trapped electron mode (TEM) to the ion temperature gradient mode (ITG mode). The plasmas with the positive correlation between the normalized density gradient and the normalized ion temperature gradient are dominated by the TEM. On the other hand, the plasmas with the negative correlation are dominated by the ITG mode