Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

Missirlian, M.; Escourbiac, F.; Merola, M.; Bobin-Vastra, I.; Schlösser, J.; Durocher, A.

doi:10.1016/j.fusengdes.2005.06.168

Cited by 15 publications

(8 citation statements)

References 2 publications

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“…The W monoblock part endured up to 1000 cycles at 10 MW/m 2 of absorbed heat flux (that is about half of the actual incident heat flux). This value is also well above, namely one order of magnitude higher, than the ITER design target for the upper part of the ITER vertical target [7].…”

Section: Development and Testing Of High Heat Flux Technologiesmentioning

confidence: 70%

“…The main outcome of this challenging and intensive R&D work was the manufacturing of a medium-scale VT prototype by the Austrian company Plansee [4], which was followed by a near full-scale VT prototype [5]. This prototype was tested in the high heat flux FE200 electron beam facility [6] operated by the company AREVA, Framatome ANP at Le Creusot in France.…”

Section: Development and Testing Of High Heat Flux Technologiesmentioning

confidence: 99%

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EU activities in preparation of the procurement of the ITER divertor

Merola¹,

Palmer²

2006

Fusion Engineering and Design

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Section: Development and Testing Of High Heat Flux Technologiesmentioning

confidence: 70%

Section: Development and Testing Of High Heat Flux Technologiesmentioning

confidence: 99%

EU activities in preparation of the procurement of the ITER divertor

Merola¹,

Palmer²

2006

Fusion Engineering and Design

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“…This is based on their excellent thermal shock resistance [1,2], absence of melting in comparison with metals and ability to retain the strength up to high temperatures [3,4]. These are desirable properties for the strike point area of the ITER divertor during the start-up phase without tritium [5][6][7] as well as for the divertor of WENDELSTEIN 7-X [8,9].…”

Section: Mechanical and Thermo-physical Characterization Of 3-directimentioning

confidence: 99%

Mechanical and thermo-physical characterization of three-directional carbon fiber composites for W-7X and ITER

Pintsuk

Compan

Koppitz

et al. 2009

Fusion Engineering and Design

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Carbon fiber composites (CFCs) are the first choice as plasma facing materials for the strike points of divertor targets for future nuclear fusion devices like WENDELSTEIN 7-X and ITER. For the application in these facilities several potential European 3D-CFCs were compared and qualified: 1) four material batches of NB31 produced by Snecma Propulsion Solide (SNECMA); 2) NB41, SNECMA, the upgraded version of NB31; 3) N31, SNECMA, which is densified by chemical vapor infiltration (CVI) instead of a final liquid pitch infiltration characteristic for NB31; 4) a new developmental 3D-CFC produced by DUNLOP.The characterization of the composites is comprised of thermo-physical measurements and tensile tests. The results are correlated to density and microstructure and summarized as follows: 1) NB41 provides the highest thermal conductivity in the ex-pitch direction of ~375 W/m.K at room temperature; 2) all material grades are, due to their heterogeneity, characterized by a relatively large scatter of mechanical properties; 3) the different densification process for N31 in comparison to NB31 has no influence on the material properties; 4) NB41 provides in all 3 directions a comparably high tensile strength with an average in the ex-pitch direction of ~220 MPa; 5) the 3D-

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“…Missiaen (2) , P. Brustolin (3) , O. Ozer (2) , A. Durocher (1) , C. Ruset (4) , C.P. Lungu (4) , X. Courtois (1) , C. Dominicy (5) , H. Maier (6) , C. Grisolia (1) , G. Piazza (7) , P Chappuis (1) (1) Euratom -CEA, Direction des sciences de matière, département de recherches sur la fusion contrôlée, CEA Cadarache, Saint Paul les Durances, France (2) LTPCM, UMR CNRS 5614, Université Joseph Fourier, Grenoble, France (3) Institut de Soudure, YUTZ, France (4) NILPRP, Magurele-Bucharest, Romania (5) CP2M, Université d'Aix Marseille III, Marseille, France (6) Max-Planck-Institut für Plasmaphysik, Euratom Association, Garching, Germany (7) EFDA JET , Close Support Unit, Culham Science Centre, UK…”

Section: Recent Developments Toward the Use Of Tungsten As Armour Matunclassified

“…The current reference concepts for tungsten armoured plasma facing components all rely on contacts bonds (hot isostatic pressing, active metal casting, brazing), with typical cross -bond dimensions of less than 0.1 mm [2][3][4]. Test mock-ups have undergone successfully cycled high heat flux tests [5], however such bonds are still considered unfavourable regarding high temperature deformation, cycle fatigue, creep or thermal shocks [6,7]. Bonds with larger dimensions (more than 1mm) and a more intimate intermixing of the two materials would be more favourable.…”

Section: Introductionmentioning

confidence: 99%

Recent developments toward the use of tungsten as armour material in plasma facing components

Mitteau¹,

Missiaen²,

Brustolin³

et al. 2007

Fusion Engineering and Design

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Future fusion experiments will rely on tungsten armour tile for their plasma facing components. In order to sustain steady state operation, the components need to be cooled through an attachment to a heat sink. All current reference concepts rely on contact bonds, unfavourable for long term application (high temperature service, cycle fatigue, thermal shocks). Three routes toward the development of thick tungsten bonds are presented here, namely functionally graded tungsten copper assembly, electron beam welding of tungsten, and coating processes. All present favourable prospects, and tend to indicate that a thick bond is possible with tungsten. Dedicated programs as well as industrial implication are however required if such concepts are to be used actually for the fabrication of large components series.

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Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

Cited by 15 publications

References 2 publications

EU activities in preparation of the procurement of the ITER divertor

EU activities in preparation of the procurement of the ITER divertor

Mechanical and thermo-physical characterization of three-directional carbon fiber composites for W-7X and ITER

Recent developments toward the use of tungsten as armour material in plasma facing components

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