On the possibility of ITER starting with full carbon

Tanabe, T.

doi:10.1016/j.fusengdes.2005.08.083

Cited by 38 publications

(21 citation statements)

References 39 publications

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“…It is interesting however to note that the extent of surface corrugation seems to change little after 60 min irradiation. This morphological change is likely to facilitate the crack-induced tritium retention during fusion operation [20][21][22]. On the other hand, the abovementioned morphological feature contrasts with earlier reports that the ion-irradiation yields diamond-like nanoparticles on the surface of HOPG.…”

Section: Characterizationscontrasting

confidence: 57%

Influence of H+ ion irradiation on the surface and microstructural changes of a nuclear graphite

Yang

Choe

Jin

et al. 2012

Fusion Engineering and Design

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Section: Characterizationscontrasting

confidence: 57%

Influence of H+ ion irradiation on the surface and microstructural changes of a nuclear graphite

Yang

Choe

Jin

et al. 2012

Fusion Engineering and Design

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“…Hence we should keep C materials as an alternative for Plasma Facing Materials (PFM) in ITER and even for a reactor. [18] …”

Section: In-vessel Tritium Inventorymentioning

confidence: 99%

Tritium issues to be solved for establishment of a fusion reactor

Tanabe

2012

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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On the possibility of ITER starting with full carbon

Cited by 38 publications

References 39 publications

Influence of H+ ion irradiation on the surface and microstructural changes of a nuclear graphite

Influence of H+ ion irradiation on the surface and microstructural changes of a nuclear graphite

Tritium issues to be solved for establishment of a fusion reactor

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

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