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Title: Heat flux mitigation by impurity seeding in high-field tokamaks

Authors:
Publication Date:
Sponsoring Org.:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
OSTI Identifier:
1339178
Grant/Contract Number:
AC05-00OR22725
Resource Type:
Journal Article: Publisher's Accepted Manuscript
Journal Name:
Nuclear Fusion
Additional Journal Information:
Journal Volume: 57; Journal Issue: 3; Related Information: CHORUS Timestamp: 2017-01-13 04:13:37; Journal ID: ISSN 0029-5515
Publisher:
IOP Publishing
Country of Publication:
IAEA
Language:
English

Citation Formats

Reinke, M. L. Heat flux mitigation by impurity seeding in high-field tokamaks. IAEA: N. p., 2017. Web. doi:10.1088/1741-4326/aa5145.
Reinke, M. L. Heat flux mitigation by impurity seeding in high-field tokamaks. IAEA. doi:10.1088/1741-4326/aa5145.
Reinke, M. L. Fri . "Heat flux mitigation by impurity seeding in high-field tokamaks". IAEA. doi:10.1088/1741-4326/aa5145.
@article{osti_1339178,
title = {Heat flux mitigation by impurity seeding in high-field tokamaks},
author = {Reinke, M. L.},
abstractNote = {},
doi = {10.1088/1741-4326/aa5145},
journal = {Nuclear Fusion},
number = 3,
volume = 57,
place = {IAEA},
year = {Fri Jan 13 00:00:00 EST 2017},
month = {Fri Jan 13 00:00:00 EST 2017}
}

Journal Article:
Free Publicly Available Full Text
Publisher's Version of Record at 10.1088/1741-4326/aa5145

Citation Metrics:
Cited by: 4works
Citation information provided by
Web of Science

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  • Cited by 5
  • Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production ofmore » submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.« less
  • Experiments conducted in high-performance 1.0 MA and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high {beta}{sub t} = 15-25%, a high bootstrap current fraction f{sub BS} = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the strongly-shaped lower single null configuration with elongation {kappa} = 2.2-2.4 and triangularity {delta} = 0.6-0.8. Divertor peak heat fluxes weremore » reduced from 6-12 MW/m{sup 2} to 0.5-2 MW/m{sup 2} in ELMy H-mode discharges using the inherently high magnetic flux expansion f{sub m} = 16-25 and the partial detachment of the outer strike point at several D{sub 2} injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Z{sub eff} were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction of ion flux to the plate, and a large neutral compression ratio. Spectroscopic measurements indicated a formation of a high-density, low temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and CII, CIII emission rates was measured.« less
  • Experiments conducted in high-performance 1.0 and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly shaped spherical torus (ST) configuration. Improved plasma performance with high beta(t) = 15-25%, a high bootstrap current fraction f(BS) = 45-50%, longer plasma pulses and an H-mode regime with smaller ELMs has been achieved in the strongly shaped lower single null configuration with elongation kappa = 2.2-2.4 and triangularity delta = 0.7-0.8. Divertor peak heat fluxes were reducedmore » from 6-12 to 0.5-2 MW m(-2) in ELMy H-mode discharges using the inherently high magnetic flux expansion f(m) = 15-25 and the partial detachment of the outer strike point at several D-2 injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Z(eff) were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction in ion flux to the plate and a large neutral compression ratio. Spectroscopic measurements indicated the formation of a high-density, low-temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and C II, CIII emission rates were measured.« less