Large Helical Device
The Large Helical Device (大型ヘリカル装置, Ōgata Herikaru Sōchi) (LHD) is a fusion research device located in Toki, Gifu, Japan. It is operated by the National Institute for Fusion Science, and is the world's second-largest superconducting stellarator, after Wendelstein 7-X. The LHD employs a heliotron magnetic field originally developed in Japan. The objective of the project is to conduct fusion plasma confinement research in a steady state in order to elucidate possible solutions to physics and engineering problems in helical plasma reactors. The LHD uses neutral beam injection, ion cyclotron radio frequency (ICRF), and electron cyclotron resonance heating (ECRH) to heat the plasma, much like conventional tokamaks. The helical divertor heat and particle exhaust system uses the large helical coils to produce a diverting field. This configuration allows for the modification of the stochastic layer size, which is positioned between the confined plasma volume and the field lines that terminate on the divertor plate. Boundary plasma research at LHD focuses on the capability of the helical divertor as an exhaust system for heliotrons and stellarators.
| Large Helical Device | |
|---|---|
 The Large Helical Device in 2014 | |
| Device type | Heliotron |
| Location | Toki, Japan |
| Affiliation | National Institute for Fusion Science |
| Technical specifications | |
| Major radius | 3.9 m (13 ft) |
| Minor radius | 0.6 m (2 ft 0 in) |
| Magnetic field | 3.0 T (30,000 G) |
| History | |
| Year(s) of operation | 1998–2025 |
The Large Helical Device (大型ヘリカル装置, Ōgata Herikaru Sōchi) (LHD) is a fusion research device located in Toki, Gifu, Japan. It is operated by the National Institute for Fusion Science, and is the world's second-largest superconducting stellarator, after Wendelstein 7-X. The LHD employs a heliotron magnetic field originally developed in Japan.
The objective of the project is to conduct fusion plasma confinement research in a steady state in order to elucidate possible solutions to physics and engineering problems in helical plasma reactors. The LHD uses neutral beam injection, ion cyclotron radio frequency (ICRF), and electron cyclotron resonance heating (ECRH) to heat the plasma, much like conventional tokamaks. The helical divertor heat and particle exhaust system uses the large helical coils to produce a diverting field. This configuration allows for the modification of the stochastic layer size, which is positioned between the confined plasma volume and the field lines that terminate on the divertor plate. Boundary plasma research at LHD focuses on the capability of the helical divertor as an exhaust system for heliotrons and stellarators.[1][2]
History
[edit]- Design finalized 1987
- Start of construction 1990
- Plasma operations from 1998
- Neutral beam injection of 3 MW was used in 1999.[3]
- In 2005 it maintained a plasma for 3,900 seconds.[4]
- In 2006 a new helium cooler was added. Using the new cooler, by 2018 a total of 10 long term operations have been achieved, reaching a maximum power level of 11.833 kA.[5]
- To aid public acceptance, an exhaust system was designed to catch and filter the radioactive tritium the fusion process produces.[6]
![[icon]](/en/wiki/File:Wiki_letter_w_cropped.svg){kind=small} | This section needs expansion. You can help by adding missing information. (March 2018) |
- The Large Helical Device experiment was concluded on Christmas Day of 2025 after more than 27 years of operation.[7]
See also
[edit]References
[edit]- ^ Morisaki, T; et al. (2013). "Initial experiments towards edge plasma control with a closed helical divertor in LHD". Nucl. Fusion. 53 (6) 063014. Bibcode:2013NucFu..53f3014M. doi:10.1088/0029-5515/53/6/063014. S2CID 122537627.
- ^ Bader, Aaron; Effenberg, Florian; Hegna, Chris C. (December 6, 2018). "Progress in Divertor and Edge Transport Research for Stellarator Plasmas" (PDF). Archived from the original (PDF) on 2023-07-26.
- ^ Fujiwara, M.; Yamada, H.; Ejiri, A.; Emoto, M.; Funaba, H.; Goto, M.; Ida, K.; Idei, H.; Inagaki, S.; Kado, S.; Kaneko, O.; Kawahata, K.; Kobuchi, T.; Komori, A.; Kubo, S.; Kumazawa, R.; Masuzaki, S.; Minami, T.; Miyazawa, J.; Morisaki, T.; Morita, S.; Murakami, S.; Muto, S.; Mutoh, T.; Nagayama, Y.; Nakamura, Y.; Nakanishi, H.; Narihara, K.; Nishimura, K.; et al. (1999). "Plasma confinement studies in LHD". Nuclear Fusion. 39 (11Y): 1659–1666. Bibcode:1999NucFu..39.1659F. doi:10.1088/0029-5515/39/11Y/305. S2CID 250824691.
Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m−3 keV s at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m−3
- ^ Achievement of One Hour Discharge with ECH on LHD 2005
- ^ Hamaguchi, Imagawa, Obana, Yanagi and Mito (2018). "Operations of the Helium Subcooling System for the LHD HelicalCoils during Ten Plasma Experimental Campaigns". Plasma and Fusion Research. 13 3405057. Bibcode:2018PFR....1305057H. doi:10.1585/pfr.13.3405057.
{{cite journal}}: CS1 maint: multiple names: authors list (link) - ^ "Design and commissioning of the exhaust detritiation system for the Large Helical Device". ResearchGate. Retrieved 2019-03-04.
- ^ 山田, 弘司. "御礼、LHD実験を完遂". 自然科学研究機構 核融合科学研究所. Retrieved 13 March 2026.
External links
[edit]- Large Helical Device Website
- Super Dense Core plasmas in LHD. Harris. 2008 16 slides. advanced - inc ballooning mode and future development options
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