Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod's high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.

doi:10.1088/1741-4326/aa7923

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod's high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

Journal Article · Wed Jul 12 00:00:00 EDT 2017 · Nuclear Fusion

DOI:https://doi.org/10.1088/1741-4326/aa7923· OSTI ID:1894446

^[1]; Wolfe, S. M. ^[2]; ^[2]; Kuang, A. Q. ^[2]; ^[3]; Reinke, M. L. ^[4]; Hubbard, A. ^[2]; Hughes, J. ^[2]; Mumgaard, R. T. ^[2]; ^[2]; Umansky, M. V. ^[5]

Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); OSTI
Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
Univ. of York (United Kingdom)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N₂) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW m^-2 down to <10 MW m^-2 while maintaining excellent core confinement, H₉₈ > 1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. Finally, these considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

View Accepted Manuscript (DOE)

Research Organization:: Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Plasma Science and Fusion Center

Sponsoring Organization:: USDOE Office of Science (SC), Fusion Energy Sciences (FES); Royal Society

Contributing Organization:: Alcator C-Mod Team

Grant/Contract Number:: FC02-99ER54512

OSTI ID:: 1894446

Alternate ID(s):: OSTI ID: 22925829

Journal Information:: Nuclear Fusion, Journal Name: Nuclear Fusion Journal Issue: 8 Vol. 57; ISSN 0029-5515

Publisher:: IOP ScienceCopyright Statement

Country of Publication:: United States

Language:: English

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Cited By (4)

The surface eroding thermocouple for fast heat flux measurement in DIII-D Ren, J.; Donovan, D.; Watkins, J. Review of Scientific Instruments, Vol. 89, Issue 10 https://doi.org/10.1063/1.5038677	journal	October 2018
Plasma detachment in divertor tokamaks Leonard, A. W. Plasma Physics and Controlled Fusion, Vol. 60, Issue 4 https://doi.org/10.1088/1361-6587/aaa7a9	journal	February 2018
Radiative heat exhaust in Alcator C-Mod I-mode plasmas Reinke, M. L.; Brunner, D.; Golfinopoulos, T. Nuclear Fusion, Vol. 59, Issue 4 https://doi.org/10.1088/1741-4326/ab04cf	journal	March 2019
Study of passively stable, fully detached divertor plasma regimes attained in innovative long-legged divertor configurations Umansky, M. V.; LaBombard, B.; Brunner, D. Nuclear Fusion, Vol. 60, Issue 1 https://doi.org/10.1088/1741-4326/ab46f4	journal	October 2019

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