Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

Journal Article · Wed Nov 16 00:00:00 EST 2016 · IEEE Transactions on Plasma Science

DOI:https://doi.org/10.1109/TPS.2016.2625325· OSTI ID:1343020

Soukhanovskii, V. A.  ^[1]; Allen, S. L. ^[1]; Fenstermacher, M. E. ^[1]; Lasnier, C. J. ^[1]; Makowski, M. A. ^[1]; McLean, A. G. ^[1]; Meier, E. T. ^[2]; Meyer, W. H. ^[1]; Rognlien, T. D. ^[1]; Ryutov, D. D. ^[1]; Scotti, F. ^[1]; Kolemen, E. ^[3]; Bell, R. E. ^[3]; Diallo, A. ^[3]; Gerhardt, S. ^[3]; Kaita, R. ^[3]; Kaye, S. ^[3]; LeBlanc, B. P. ^[3]; Maingi, R. ^[3]; Menard, J. E. ^[3] ; Podesta, M. ^[3]; Roquemore, A. L. ^[3]; Groebner, R. J. ^[4]; Hyatt, A. W. ^[4]; Leonard, A. W. ^[4]; Osborne, T. H. ^[4]; Petrie, T. W. ^[4]; Ahn, J. -W. ^[5]; Raman, R. ^[6]; Watkins, J. G. ^[7]

---

1. Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2. Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); College of William and Mary, Williamsburg, VA (United States)
3. Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
4. General Atomics, San Diego, CA (United States)
5. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
6. Univ. of Washington, Seattle, WA (United States)
7. Sandia National Lab. (SNL-CA), Livermore, CA (United States)

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D₂-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D₂-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q_peak reduction factors (see standard divertor).

• Research Organization: Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
• Sponsoring Organization: USDOE
• Grant/Contract Number: AC52-07NA27344
• OSTI ID: 1343020
• Report Number(s): LLNL-JRNL-675395; TRN: US1701246
• Journal Information: IEEE Transactions on Plasma Science, Vol. 44, Issue 12; ISSN 0093-3813
• Publisher: IEEE
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION
tokamaks
divertors
plasma materials interactions