ITER R&amp;D: Vacuum Vessel and In-Vessel Components: Divertor Cassette

Tivey, R.; Akiba, Masato; Mazul, I.; Merola, M.; Ulrickson, M.

doi:10.1016/s0920-3796(01)00212-5

Cited by 59 publications

(20 citation statements)

References 7 publications

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“…The current status of R&D on plasma facing materials and components are comprehensively summarized in recent review papers. [5][6][7][8][9] The followings are a brief summary of the currently promising options for ITER PFCs and non-destructive examinations applied during manufacturing.…”

Section: Plasma Facing Materials and Components In Itermentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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Section: Plasma Facing Materials and Components In Itermentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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“…In the present study, WCHFs were compared with Tong-75 correlation [14], which was originally developed for the light-water fission reactor. This correlation has been extensively used to estimate WCHF and also applied in the design analysis of ITER divertor [1]. In each condition, the experimental results have higher WCHF than that of the predicted values by Tong-75 correlation, which shows the same tendency as the screw tube has higher ICHF than that of a smooth tube in the previous results [7].…”

Section: Critical Heat Flux At Higher Pressure and Temperaturementioning

confidence: 67%

“…In particular, the surface heat flux onto divertor high heat flux components is estimated to be 10-20 MW/m 2 in ITER [1]. In the design of such high heat load removal components, we have to consider several issues affecting the performance and lifetime of PFCs and influencing the choice of cooling methods and materials.…”

Section: Introductionmentioning

confidence: 99%

Experimental examination of heat removal limitation of screw cooling tube at high pressure and temperature conditions

Ezato

Suzuki

Dairaku

et al. 2006

Fusion Engineering and Design

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“…In the EDA, various elements for high heat flux components were developed and fabricated by the four parties. High heat cycle tests showed that carbon fiber composite (CFC) monoblock survived 2000 cycles at 20 MW and the tungsten armors survived 15 MWm −2 for 1000 cycles without degradation of the thermal performance [16][17][18]. The structural housing named "divertor cassette" was also fabricated with a required tolerance and was confirmed to withstand the thermal and mechanical loads to be imposed during ITER normal operation and transients.…”

Section: Divertormentioning

confidence: 99%

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

Matsuda

Tobita²

2013

Journal of Nuclear Science and Technology

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An international joint project of fusion experimental reactor, the ITER (International Thermonuclear Experimental Reactor), is reviewed in view of long-range fusion energy research and development (R&D). Its purpose, goal, evolution, and the present construction status are briefly reviewed. While the ITER is a core machine in the present stage, generation of electricity is a role of the next-step fusion demonstration power plant "DEMO." The status of designs and technology R&D for DEMO are also reviewed.

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ITER R&D: Vacuum Vessel and In-Vessel Components: Divertor Cassette

Cited by 59 publications

References 7 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Experimental examination of heat removal limitation of screw cooling tube at high pressure and temperature conditions

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

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