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Title: Diagnostic options for radiative divertor feedback control on NSTX-U

Abstract

A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q{sub peak} Less-Than-Or-Slanted-Equal-To 15 MW/m{sup 2}), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D{sub 2} or CD{sub 4} gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m{sup 2}, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic 'security' monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

Authors:
;  [1]; ;  [2];  [3]
  1. Lawrence Livermore National Laboratory, Livermore, California, 94550 (United States)
  2. Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)
  3. University of Washington, Seattle, Washington 98195 (United States)
Publication Date:
OSTI Identifier:
22093858
Resource Type:
Journal Article
Journal Name:
Review of Scientific Instruments
Additional Journal Information:
Journal Volume: 83; Journal Issue: 10; Other Information: (c) 2012 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA); Journal ID: ISSN 0034-6748
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; 46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; CONTROL SYSTEMS; DAMAGE; DEUTERIUM; DIVERTORS; FEEDBACK; FIRST WALL; GAS INJECTION; GRAPHITE; HEAT FLUX; HEATING LOAD; H-MODE PLASMA CONFINEMENT; IMPURITIES; INFRARED THERMOGRAPHY; ITER TOKAMAK; NSTX DEVICE; PLASMA; PLASMA DIAGNOSTICS; PLASMA SCRAPE-OFF LAYER; RECOMBINATION; SPHERICAL CONFIGURATION; STEADY-STATE CONDITIONS; SURFACES

Citation Formats

Soukhanovskii, V. A., McLean, A. G., Gerhardt, S. P., Kaita, R., and Raman, R. Diagnostic options for radiative divertor feedback control on NSTX-U. United States: N. p., 2012. Web. doi:10.1063/1.4732176.
Soukhanovskii, V. A., McLean, A. G., Gerhardt, S. P., Kaita, R., & Raman, R. Diagnostic options for radiative divertor feedback control on NSTX-U. United States. doi:10.1063/1.4732176.
Soukhanovskii, V. A., McLean, A. G., Gerhardt, S. P., Kaita, R., and Raman, R. Mon . "Diagnostic options for radiative divertor feedback control on NSTX-U". United States. doi:10.1063/1.4732176.
@article{osti_22093858,
title = {Diagnostic options for radiative divertor feedback control on NSTX-U},
author = {Soukhanovskii, V. A. and McLean, A. G. and Gerhardt, S. P. and Kaita, R. and Raman, R.},
abstractNote = {A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q{sub peak} Less-Than-Or-Slanted-Equal-To 15 MW/m{sup 2}), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D{sub 2} or CD{sub 4} gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m{sup 2}, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic 'security' monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).},
doi = {10.1063/1.4732176},
journal = {Review of Scientific Instruments},
issn = {0034-6748},
number = 10,
volume = 83,
place = {United States},
year = {2012},
month = {10}
}