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Journal Articles

Neutron shielding and blanket neutronics design

Yamauchi, Michinori*; Nishitani, Takeo; Nishio, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 80(11), p.952 - 954, 2004/11

Considering the geometrical characteristics of tokamak reactors with low aspect ratio, a basic neutronics strategy was derived to construct the inboard structure mainly for neutron shielding and produce enough tritium in the outboard blanket. The designs for optimal inboard shield were surveyed and necessary thickness was estimated to make the neutron flux low enough on the super-conducting magnet. In addition, the outer blanket designs were studied to attain the tritium breeding ratio (TBR) large enough for a self-sustaining fusion reactor on the basis of the advanced fusion reactor materials.

JAEA Reports

Integrity of the first wall in fusion reactors

Kurihara, Ryoichi

JAERI-Tech 2004-052, 39 Pages, 2004/07

JAERI-Tech-2004-052.pdf:2.1MB

The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel.

Journal Articles

Fracture mechanics evaluation of a crack generated in SiC/SiC composite first wall

Kurihara, Ryoichi; Ueda, Shuzo; Nishio, Satoshi; Seki, Yasushi

Fusion Engineering and Design, 54(3-4), p.465 - 471, 2001/04

 Times Cited Count:10 Percentile:58.96(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Forced convective heat transfer in square-ribbed coolant channels with helium gas for fusion power reactors

Takase, Kazuyuki

Fusion Engineering and Design, 49-50, p.349 - 354, 2000/11

 Times Cited Count:6 Percentile:42.55(Nuclear Science & Technology)

no abstracts in English

Journal Articles

A Fusion power reactor concept using SiC/SiC composites

Ueda, Shuzo; Nishio, Satoshi; Seki, Yasushi; Kurihara, Ryoichi; Adachi, Junichi*; Yamazaki, Seiichiro*; DREAM-Design-Team

Journal of Nuclear Materials, 258-263, p.1589 - 1593, 1998/00

 Times Cited Count:43 Percentile:93.84(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

JAEA Reports

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