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Title: Thermal-hydraulics modeling for prototype testing of the W7-X high heat flux scraper element

The long-pulse operation of the Wendelstein 7-X (W7-X) stellarator experiment is scheduled to begin in 2020. This operational phase will be equipped with water-cooled plasma facing components to allow for longer pulse durations. Certain simulated plasma scenarios have been shown to produce heat fluxes that surpass the technological limits on the edges of the divertor target elements during steady-state operation. In order to reduce the heat load on the target elements, the addition of a “scraper element” (SE) is under investigation. The SE is composed of 24 water-cooled carbon fiber reinforced carbon composite monoblock units. Multiple full-scale prototypes have been tested in the GLADIS high heat flux test facility. Previous computational studies revealed discrepancies between the simulations and experimental measurements. In this work, single-phase thermal-hydraulics modeling was performed in ANSYS CFX to identify potential causes for such discrepancies. Possible explanations investigated were the effects of a non-uniform thermal contact resistance and a potential misalignment of the monoblock fibers. And while the difference between the experimental and computational results was not resolved by a non-uniform thermal contact resistance, the computational results provided insight into the potential performance of a W7-X monoblock unit. Circumferential temperature distributions highlighted the expected boiling regions ofmore » such a unit. Finally, simulations revealed that modest angles of fiber misalignment in the monoblocks result in asymmetries at the unit edges and provide temperature differences similar to the experimental results.« less
 [1] ; ORCiD logo [2] ;  [3] ;  [3] ;  [1]
  1. Univ. of Tennessee, Knoxville, TN (United States)
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  3. Max Planck Inst. for Plasma Physics, Garching (Germany)
Publication Date:
Grant/Contract Number:
Accepted Manuscript
Journal Name:
Fusion Engineering and Design
Additional Journal Information:
Journal Volume: 121; Journal Issue: C; Journal ID: ISSN 0920-3796
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
Country of Publication:
United States
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; plasma facing components; monoblock; CFC; high heat flux
OSTI Identifier: