ITER divertor, design issues and research and development

Tivey, R.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Janeschitz, G.; Raffray, R.; Akiba, Masato; Mazul, I.; Pacher, H.D.; Ulrickson, M.; Vieider, G.

doi:10.1016/s0920-3796(99)00047-2

Cited by 42 publications

(15 citation statements)

References 7 publications

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“…doi:10.1016/j.jnucmat.2004.07.054 thermal properties, low solubility for hydrogen and low sputtering yield. In fact, tungsten has been selected as a plasma facing material in the divertor region of ITER [2]. There are long lasting attempts of plasma irradiation, e.g., in ASDEX Upgrade [3].…”

Section: Introductionmentioning

confidence: 99%

Microscopic and macroscopic damage in metals exposed to LHD divertor plasmas

Tokitani

Miyamoto

Koga

et al. 2005

Journal of Nuclear Materials

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Section: Introductionmentioning

confidence: 99%

Microscopic and macroscopic damage in metals exposed to LHD divertor plasmas

Tokitani

Miyamoto

Koga

et al. 2005

Journal of Nuclear Materials

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“…1, the preliminary divertor design for FDS-III is carried out and the basic structure, shown in Fig. 4, similar to that of ITER using ''vertical targets" [16] mainly consists of inner target plate, outer target, dome, baffles, and cassettes. The oxide dispersion strengthened (ODS) RAFM steel is used as structure and LiPb as coolant.…”

Section: Divertormentioning

confidence: 99%

Fusion-based hydrogen production reactor and its material selection

Team

2009

Journal of Nuclear Materials

118

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“…3), and placed between the existing in-vessel components (IVCs) and the plasma volume in an extremely limited space. CFC monoblocks designed for ITER have been qualified to survive 20 MW/m 2 heat flux [4][5][6]. The convective heat fluxes on the divertor components have been modeled using magnetic field configurations where the contribution due to equilibrium plasma currents has been included [7].…”

Section: Introductionmentioning

confidence: 99%

Modeling and Analysis of the W7-X High Heat-Flux Divertor Scraper Element

Lumsdaine

Boscary

Clark

et al. 2014

IEEE Trans. Plasma Sci.

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Abstract-The Wendelstein 7-X stellarator experiment is scheduled for the completion of device commissioning and the start of first plasma in 2015. At the completion of the first two operational phases, the inertially cooled test divertor unit will be replaced with an actively cooled high heat-flux divertor which will enable the device to increase its pulse length to steady-state plasma performance.Plasma simulations show that the evolution of bootstrap current in certain plasma scenarios produce excessive heat fluxes on the edge of the divertor targets. It is proposed to place an additional "scraper element" in the ten divertor locations to intercept some of the plasma flux and reduce the heat load on these divertor edge elements. Each scraper element may experience a 500 kW steady-state power load, with localized heat fluxes as high as 20 MW/m 2 . Computational analysis has been performed in order to examine the thermal integrity of the scraper element.The peak temperature in the CFC, the total pressure drop in the cooling water, and the increase in water temperature must all be examined to stay within specific design limits. Computational fluid dynamics (CFD) modeling is performed to examine the flow paths through the multiple monoblock fingers as well as the thermal transfer through the monoblock swirl tube channels.

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ITER divertor, design issues and research and development

Cited by 42 publications

References 7 publications

Microscopic and macroscopic damage in metals exposed to LHD divertor plasmas

Microscopic and macroscopic damage in metals exposed to LHD divertor plasmas

Fusion-based hydrogen production reactor and its material selection

Modeling and Analysis of the W7-X High Heat-Flux Divertor Scraper Element

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