Plasma-Material Interface Development for Future Spherical Tokamak-based Devices in NSTX.

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The divertor plasma-material interface (PMI) must be able to withstand steady-state heat fluxes up to 10 MW/m{sup 2} (a limit imposed by the present day divertor material and engineering constraints) with minimal material erosion, as well as to provide impurity control and ion density pumping capabilities. In spherical tokamaks (STs), the compact divertor geometry and the requirement of low core electron collisionality n*{sub e} at n{sub e} < 0.5-0.7 n{sub G} (where n{sub G} is the Greenwald density) for increased neutral beam current drive efficiency impose much greater demands on divertor and first-wall particle and heat flux mitigation solutions. In ... continued below

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Soukhanovskii, V. A.; Battaglia, D.; Bell, M G.; Bell, R. E.; Diallo, A.; Gerhardt, S. et al. September 24, 2011.

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The divertor plasma-material interface (PMI) must be able to withstand steady-state heat fluxes up to 10 MW/m{sup 2} (a limit imposed by the present day divertor material and engineering constraints) with minimal material erosion, as well as to provide impurity control and ion density pumping capabilities. In spherical tokamaks (STs), the compact divertor geometry and the requirement of low core electron collisionality n*{sub e} at n{sub e} < 0.5-0.7 n{sub G} (where n{sub G} is the Greenwald density) for increased neutral beam current drive efficiency impose much greater demands on divertor and first-wall particle and heat flux mitigation solutions. In NSTX, divertor heat flux mitigation and impurity control with an innovative 'snowflake' divertor configuration and ion density pumping by evaporated lithium wall and divertor coatings are studied. Lithium coatings have enabled ion density reduction up to 50% in NSTX through the reduction of wall and divertor recycling rates. The 'snowflake' divertor configuration was obtained in NSTX in 0.8-1 MA 4-6 MW NBI-heated H-mode lithium-assisted discharges using three divertor coils. The snowflake divertor formation was always accompanied by a partial detachment of the outer strike point with an up to 50% increase in divertor radiation from intrinsic carbon, the peak divertor heat flux reduction from 3-6 MW/m{sup 2} to 0.5-1 MW/m{sup 2}, and a significant increase in divertor volume recombination. High core confinement was maintained with the snowflake divertor, evidenced by the t{sub E}, W{sub MHD} and the H98(y,2) factors similar to those of the standard divertor discharges. Core carbon concentration and radiated power were reduced by 30-70%, apparently as a result of reduced divertor physical and chemical sputtering in the snowflake divertor and ELMs. In the SFD discharges, the MHD stability of the H-mode pedestal region was altered leading to the re-appearance of medium size (DW/W = 5-10%), Type I, ELMs otherwise suppressed due to lithium conditioning. Fast divertor measurements showed that impulsive particle and heat fluxes due to the ELMs were significantly dissipated in the high magnetic flux expansion region of the snowflake divertor. The snowflake divertor configuration is being combined in experiments with extrinsic deuterium or impurity gas puffing for increased dissipative divertor power losses, additional upper divertor nulls for increased power sharing between the upper and the lower divertors, and lithium coated plasma facing components for large area ion pumping. These efforts are aimed at the development of an integrated PMI for future ST-based devices for fusion development applications.

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PDF-file: 30 pages; size: 10.5 Mbytes

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  • Presented at: The Joint Meeting of 5th IAEA Technical Meeting on Spherical Tori, 16th International Workshop on Spherical Torus (ISTW2011) and 2011 US-Japan Workshop on ST Plasma, Toki, Japan, Sep 27 - Sep 30, 2011

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  • Report No.: LLNL-CONF-501060
  • Grant Number: W-7405-ENG-48
  • Office of Scientific & Technical Information Report Number: 1035962
  • Archival Resource Key: ark:/67531/metadc833737

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  • September 24, 2011

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  • May 19, 2016, 3:16 p.m.

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  • Dec. 6, 2016, 4:17 p.m.

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Soukhanovskii, V. A.; Battaglia, D.; Bell, M G.; Bell, R. E.; Diallo, A.; Gerhardt, S. et al. Plasma-Material Interface Development for Future Spherical Tokamak-based Devices in NSTX., presentation, September 24, 2011; Livermore, California. (digital.library.unt.edu/ark:/67531/metadc833737/: accessed August 19, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.