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Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

Abstract

Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering  More>>
Authors:
Yanagi, N.; Sagara, A.; Goto, T.; Masuzaki, S.; Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp [1] 
  1. National Institute for Fusion Science, Toki (Japan)
Publication Date:
Sep 15, 2012
Product Type:
Conference
Report Number:
IAEA-CN-197; FTP/P7-37
Resource Relation:
Conference: FEC 2012: 24. IAEA Fusion Energy Conference, San Diego, CA (United States), 8-13 Oct 2012; Related Information: In: 24. IAEA Fusion Energy Conference. Programme and Book of Abstracts| 789 p.
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; ASYMMETRY; BARIUM OXIDES; BREEDING BLANKETS; COPPER OXIDES; DIVERTORS; HEAT FLUX; HELIOTRON; HIGH-TC SUPERCONDUCTORS; LHD DEVICE; MAGNETIC FIELDS; MAGNETIC SURFACES; NEUTRONS; PEAKS; PERTURBATION THEORY; PLASMA; PLASMA CONFINEMENT; STEADY-STATE CONDITIONS; YTTRIUM OXIDES
OSTI ID:
22192641
Research Organizations:
International Atomic Energy Agency, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
TRN: XA14S0051017130
Availability:
Available from INIS in electronic form. Also available on-line: http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2012/cn197/cn197_Programme.pdf
Submitting Site:
INIS
Size:
page(s) 508
Announcement Date:
Feb 20, 2014

Citation Formats

Yanagi, N., Sagara, A., Goto, T., Masuzaki, S., and Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor. IAEA: N. p., 2012. Web.
Yanagi, N., Sagara, A., Goto, T., Masuzaki, S., & Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor. IAEA.
Yanagi, N., Sagara, A., Goto, T., Masuzaki, S., and Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp. 2012. "Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor." IAEA.
@misc{etde_22192641,
title = {Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor}
author = {Yanagi, N., Sagara, A., Goto, T., Masuzaki, S., and Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp}
abstractNote = {Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering the core plasma. Though this scheme is effective in dispersing the heat flux in the poloidal direction, the toroidal asymmetry still remains. The AC operation may also give unforeseen engineering difficulties. We here propose that the peak heat flux be mitigated using RMP fields in steady-state. The magnetic field-lines are numerically traced in the vacuum configuration and their footprints coming to the divertor regions are counted. Their fraction plotted as a function of the toroidal angle indicates that the peak heat flux be mitigated to {approx} 20 MW per square meters at 3 GW fusion power generation without having radiation dispersion when an RMP field is applied. We note that the magnetic surfaces at the core region are not significantly affected by the RMP field. The poloidal sweeping by HD coils (with quasi-steady-state AC operation) should mitigate the erosion of divertor plates. We propose that these coils be fabricated using YBCO high-temperature superconductors operated at > 20K. (author)}
place = {IAEA}
year = {2012}
month = {Sep}
}