High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

T. K. Gray, D. L. Youchison, R. E. Ellis, M. A. Jaworski, A. Khodak, T. Looby, M. L. Reinke, G. Smalley, D. E. Wolfe

Research output: Contribution to journalArticlepeer-review

1 Scopus citations

Abstract

As part of the recovery project of the National Spherical Tokamak Experiment–Upgrade (NSTX-U), the divertor plasma-facing components (PFCs) were redesigned to handle significantly higher heat fluxes and longer pulse lengths than NSTX. The design process resulted in a castellated, graphite PFC tile. To verify the thermal performance of this design, dedicated electron beam, high heat flux (HHF) testing was carried out on a de-optimized mock-up PFC target. These tests demonstrated that the tile design is itself robust to large, localized thermal gradients. No mechanical damage to the mock-up was observed during HHF testing, though the actual PFC tile mechanical tie-down was not tested. Rather, when the surface temperature exceeded the sublimation temperature of graphite, carbon blooms from the mock-up tile surface were observed. This resulted in 1 to 2 mm of surface material ablating from the mock-up after repeated, highly localized electron beam exposures.

Original languageEnglish
Pages (from-to)9-18
Number of pages10
JournalFusion Science and Technology
Volume77
Issue number1
DOIs
StatePublished - 2021

Funding

This paper has been authored by a contractor of the U.S. government under contracts DE-AC05-00OR22725, DE-AC52-07NA27344, and DE-AC02-09CH11466. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Fusion Energy Sciences Program. Accordingly, the U.S. government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. government purposes.

Keywords

  • Plasma-facing component
  • graphite
  • high heat flux testing

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