Simulations of tokamak boundary plasma turbulence transport in setting the divertor heat flux width
[3];
[2];
[1];
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dalian Univ. of Technology (China)
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Peking Univ., Beijing (China); General Atomics, San Diego, CA (United States)
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Univ. of Science and Technology of China, Hefei (China)
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics
- Univ. of Science and Technology of China, Hefei (China); General Atomics, San Diego, CA (United States)
The BOUT++ code has been used to simulate edge plasma electromagnetic (EM) turbulence and transport, and to study the role of EM turbulence in setting the scrape-off layer (SOL) heat flux width λq. More than a dozen tokamak discharges from C-Mod, DIII-D, EAST, ITER and CFETR have been simulated with encouraging success. The parallel electron heat fluxes onto the target from the BOUT++ simulations of C-Mod, DIII-D, and EAST follow the experimental heat flux width scaling of the inverse dependence on the poloidal magnetic field. Further turbulence statistics analysis shows that the blobs are generated near the pedestal pressure peak gradient region inside the separatrix and contribute to the transport of the particle and heat in the SOL region. Transport simulations indicate two distinct regimes: drift dominant regime and turbulence dominant regime. Goldston's heuristic drift-based (HD) model yields a consistent divertor heat flux width in the drift dominant regime. For C-Mod enhanced Dα H-mode discharges, drifts and turbulence are competing in setting the divertor heat flux width, possibly due to its compact machine size and good pedestal confinement. The simulations for ITER and CFETR indicate that divertor heat flux width λq of the future machines may no longer follows the 1/Bpol,OMP HD-based empirical (Eich) scalings and the HD model gives a pessimistic limit of divertor heat flux width. The simulation results show a transition from a drift dominant regime to a turbulence dominant regime from current machines to future machines such as ITER and CFETR for two reasons. (1) The magnetic drift-based radial transport decreases due to large CFETR and ITER machine sizes. (2) The SOL turbulence thermal diffusivity increases due to larger turbulent fluxes ejected from the pedestal into the SOL when operating in a different pedestal structure, from an ELM-free H-mode pedestal regime to a small and grassy ELM regime.
- Research Organization:
- Lawrence Livermore National Laboratory (LLNL), Livermore, CA (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA); China Scholarship Committee
- Grant/Contract Number:
- AC52-07NA27344; 201606060097; 201706010039; 201706340070
- OSTI ID:
- 1872686
- Report Number(s):
- LLNL-JRNL-765641; 955085; TRN: US2306901
- Journal Information:
- Nuclear Fusion, Vol. 59, Issue 12; ISSN 0029-5515
- Publisher:
- IOP ScienceCopyright Statement
- Country of Publication:
- United States
- Language:
- English
Simulations of divertor heat flux width using transport code with cross-field drifts under the BOUT++ framework
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journal | January 2020 |
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