High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

Gray, Travis; Youchison, Dennis; Ellis, Robert; Jaworski, M. A.; Khodak, Andrei; Looby, Thomas; Reinke, Matthew; Smalley, Gustav; Wolfe, Douglas

doi:10.1080/15361055.2020.1831872

Title: High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

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. Finally, this resulted in 1 to 2 mm of surface material ablating from the mock-up after repeated, highly localized electron beam exposures.

Authors:

^[1];

^[1]; Ellis, Robert ^[2]; Jaworski, M. A. ^[3]; Khodak, Andrei ^[2]; Looby, Thomas ^[4];

^[1]; Smalley, Gustav ^[2]; Wolfe, Douglas ^[5]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Univ. of Tennessee, Knoxville, TN (United States)
Pennsylvania State Univ., University Park, PA (United States)

Publication Date:: Tue Nov 03 00:00:00 EST 2020

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE Office of Science (SC), Fusion Energy Sciences (FES)

OSTI Identifier:: 1831701

Grant/Contract Number:: AC05-00OR22725; AC52-07NA27344; AC02-09CH11466

Resource Type:: Accepted Manuscript

Journal Name:: Fusion Science and Technology

Additional Journal Information:: Journal Volume: 77; Journal Issue: 1; Journal ID: ISSN 1536-1055

Publisher:: American Nuclear Society

Country of Publication:: United States

Language:: English

Subject:: 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; plasma-facing component; graphite; high heat flux testing

Citation Formats


                    Gray, Travis, Youchison, Dennis, Ellis, Robert, Jaworski, M. A., Khodak, Andrei, Looby, Thomas, Reinke, Matthew, Smalley, Gustav, and Wolfe, Douglas. High Heat Flux Testing of Castellated Graphite Plasma- Facing Components.  United States: N. p., 2020. 
Web.  doi:10.1080/15361055.2020.1831872.

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                    Gray, Travis, Youchison, Dennis, Ellis, Robert, Jaworski, M. A., Khodak, Andrei, Looby, Thomas, Reinke, Matthew, Smalley, Gustav, & Wolfe, Douglas. High Heat Flux Testing of Castellated Graphite Plasma- Facing Components.  United States.  https://doi.org/10.1080/15361055.2020.1831872

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                    Gray, Travis, Youchison, Dennis, Ellis, Robert, Jaworski, M. A., Khodak, Andrei, Looby, Thomas, Reinke, Matthew, Smalley, Gustav, and Wolfe, Douglas. Tue .  
"High Heat Flux Testing of Castellated Graphite Plasma- Facing Components".  United States.  https://doi.org/10.1080/15361055.2020.1831872.  https://www.osti.gov/servlets/purl/1831701.

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@article{osti_1831701,

  title        = {High Heat Flux Testing of Castellated Graphite Plasma- Facing Components},

  author       = {Gray, Travis and Youchison, Dennis and Ellis, Robert and Jaworski, M. A. and Khodak, Andrei and Looby, Thomas and Reinke, Matthew and Smalley, Gustav and Wolfe, Douglas},

  abstractNote = {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. Finally, this resulted in 1 to 2 mm of surface material ablating from the mock-up after repeated, highly localized electron beam exposures.},

  doi          = {10.1080/15361055.2020.1831872},

  journal      = {Fusion Science and Technology},

  number       = 1,

  volume       = 77,

  place        = {United States},

  year         = {Tue Nov 03 00:00:00 EST 2020},

  month        = {Tue Nov 03 00:00:00 EST 2020}

}

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Works referenced in this record:

Plasma wall interaction and its implication in an all tungsten divertor tokamak
journal, November 2007

Neu, R.; Balden, M.; Bobkov, V.
Plasma Physics and Controlled Fusion, Vol. 49, Issue 12B
DOI: 10.1088/0741-3335/49/12B/S04

Integrated plasma facing component calorimetry for measurement of shot integrated deposited energy in the NSTX-U
journal, October 2018

Gray, T. K.; Allen, N.; Reinke, M. L.
Review of Scientific Instruments, Vol. 89, Issue 10
DOI: 10.1063/1.5039337

Physics basis for the first ITER tungsten divertor
journal, August 2019

Pitts, R. A.; Bonnin, X.; Escourbiac, F.
Nuclear Materials and Energy, Vol. 20
DOI: 10.1016/j.nme.2019.100696

Dependence of divertor heat flux widths on heating power, flux expansion, and plasma current in the NSTX
journal, August 2011

Gray, T. K.; Maingi, R.; Soukhanovskii, V. A.
Journal of Nuclear Materials, Vol. 415, Issue 1
DOI: 10.1016/j.jnucmat.2011.01.029

Graphite Tile Thermal Performance on the New DIII-D Lower Divertor
journal, October 2007

Murphy, C. J.; Anderson, P. M.; Lasnier, C. J.
Fusion Science and Technology, Vol. 52, Issue 3
DOI: 10.13182/FST07-A1544

An ITER-like wall for JET
journal, June 2007

Paméla, J.; Matthews, G. F.; Philipps, V.
Journal of Nuclear Materials, Vol. 363-365
DOI: 10.1016/j.jnucmat.2006.12.056

Exploration of spherical torus physics in the NSTX device
journal, March 2000

Ono, M.; Kaye, S. M.; Peng, Y. -K. M.
Nuclear Fusion, Vol. 40, Issue 3Y
DOI: 10.1088/0029-5515/40/3Y/316

Overview of the physics and engineering design of NSTX upgrade
journal, July 2012

Menard, J. E.; Gerhardt, S.; Bell, M.
Nuclear Fusion, Vol. 52, Issue 8
DOI: 10.1088/0029-5515/52/8/083015

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Title: High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

Abstract

Citation Formats

Plasma wall interaction and its implication in an all tungsten divertor tokamak journal, November 2007

Integrated plasma facing component calorimetry for measurement of shot integrated deposited energy in the NSTX-U journal, October 2018

Physics basis for the first ITER tungsten divertor journal, August 2019

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Exploration of spherical torus physics in the NSTX device journal, March 2000

Overview of the physics and engineering design of NSTX upgrade journal, July 2012

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journal, November 2007

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journal, October 2018

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journal, August 2019

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journal, August 2011

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journal, June 2007

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