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

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

All Science Journal Classification (ASJC) codes

  • Civil and Structural Engineering
  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • Materials Science(all)
  • Mechanical Engineering

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