The divertor plasma characteristics in the Large Helical Device

Masuzaki, S.; Morisaki, T.; Ohyabu, N.; Komori, A.; Suzuki, H.; Noda, N.; Kubota, Yoshinobu; Sakamoto, R.; Narihara, K.; Kawahata, K.; Tanaka, K.; Tokuzawa, T.; Morita, S.; Goto, M.; Osakabe, M.; Watanabe, Tsuguhiro; Matsumoto, Yutaka; Motojima, O.

doi:10.1088/0029-5515/42/6/313

Cited by 130 publications

(140 citation statements)

References 28 publications

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“…The increase in the atom flux at the inboard-side Xpoints implies a high particle recycling in the inboard-side divertor region. The Langmuir probe measurement shows high ion fluxes onto the inboard-side divertor plates [7]. This result tends to be consistent with the present results.…”

Section: T E and N E Evaluation And Inward Neutral Fluxsupporting

confidence: 93%

Poloidal Distribution Measurement of Inward Neutral Flux in LHD

Goto

Morita

2007

Plasma and Fusion Research

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Three emission lines of neutral helium, i.e., the λ 667.8 nm (2 1 P-3 1 D), λ 728.1 nm (2 1 P-3 1 S), and λ 706.5 nm (2 3 P-3 3 S) lines, are observed with an array of parallel lines-of-sight which covers an entire poloidal cross section of the plasma in the Large Helical Device (LHD). Their emission locations and intensities are determined from the Zeeman profiles. The electron temperature and density are evaluated from intensity ratios among the observed emission lines at each emission location and the poloidal distribution of ionization flux or its equivalent quantity, inward neutral particle flux, is derived with the help of collisional-radiative model calculations. A relatively strong neutral flux is observed at the inboard-side X-point and this result implies a high particle recycling in the inboardside divertor region. The result is consistent with the ion flux distribution onto the helically located divertor plates measured with Langmuir probes.

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Section: T E and N E Evaluation And Inward Neutral Fluxsupporting

confidence: 93%

Poloidal Distribution Measurement of Inward Neutral Flux in LHD

Goto

Morita

2007

Plasma and Fusion Research

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“…During the operation, the magnetic axis radius went and returned 18.5 times between 3.67 m and 3.7 m. In the heliotron configuration, the heat flux density along the divertor traces changes greatly for small changes of the axis radius near 3.65-3.7 m [8,9]. The temperatures of two different divertor plates are also shown.…”

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confidence: 94%

Thirty-Minute Plasma Sustainment by ICRF, EC and NBI Heating in the Large Helical Device

Mutoh

Kumazawa

Seki

et al. 2005

J. Plasma Fusion Res.

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Steady-state plasma heating was successfully performed and sustained for more than 30 min in the LHD. By using ICRF heating and additional EC and NBI heating, a total input energy of 1.3 GJ was achieved. The average input power was 680 kW and the plasma duration was 31 min 45 sec. The hardware of the ICRF and divertor plates was much improved and the position of the ICRF antenna was optimized. The heat load to the divertor plates was effectively dispersed by the magnetic axis swing technique, which caused large changes in the heat load distribution along the divertor leg traces. Keywords:Large Helical Device, Steady state operation, ICRF heating, Divertor, 1.3 GJ In December 2004, steady-state plasma sustainment was successfully performed for more than 30 min in the Large Helical Device (LHD) [1]. The average input power was 680 kW (ICRF [2-4] 520 kW, ECH 100 kW and NBI 60 kW) and the plasma duration was 31 min 45 sec. Before the 2003 campaign, the plasma sustainment time of the ICRF longpulse experiment was limited by local temperature rises of the divertor carbon plates near the ICRF antenna section, and gradual increase of out-gassing from the wall finally terminated the plasma operation [5]. After that experiment, a new mechanical structure and new carbon material sheets were used to suppress the temperature rise and the out-gassing rate. The hardware of ICRF heating was also much improved.The steady-state operation was performed using the standard experimental configuration of the LHD. The magnetic axis radius was around 3.67 to 3.7 m, the magnetic strength was 2.75 T at 3.6 m, and the pitch parameter of the helical winding was 1.254. The helical coil current was fixed and the vertical coil currents were controlled. Three ICRF antennas were used and the frequency was 38.47 MHz. In author's e-mail: mutoh@nifs.ac.jp this condition, the cyclotron resonance region of minority ions is located near the saddle point on the mod-B contour plane in the cross-section of the plasma. In this mode, the ICRF wave power heated minority ions and did not heat electrons directly. The LHD vacuum chamber was conditioned by boronization, which seemed to be effective in suppressing the impurity influx to low levels during steady-state operations.The plasma parameters are shown in Fig. 1. The ICRF and ECH were continuously injected, and NBI was repetitively injected by the 25 sec pulse operations. The plasma was mainly sustained by ICRF and partially supported by ECH [6] and NBI [7]. The ECH and NBI helped to control some disturbances, for example, occasional dust particles from the upper vacuum vessel. During the operation time of 31 min, helium gas was fed by a puffing system to form the majority ion species for the ICRF heating. The repetitive NBI pulses worked to heat the plasma and also to keep the hydrogen concentration ratio within a suitable range for the minority-heating mode of ICRF. Without the NBI pulses, the ratio of Hα to Helium I lines gradually decreased by a factor

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“…The field lines in the ergodic layer are connected to the divertor plates through the edge surface layer. The connection length of the open field lines in the ergodic layer are over several hundreds meters while the field line length from the X-point to the divertor plates is a few meters for the strong poloidal component of magnetic field in the divertor region [2,3].…”

Section: Introductionmentioning

confidence: 99%

“…It was revealed that particle flux position and deposition profile are mainly determined by the field lines structure on the plates [3]. However the detailed analysis of the particle flux profiles on the divertor plates has not been conducted.…”

Section: Introductionmentioning

confidence: 99%

Investigation on the influence of plasma properties and SOL transport on the particle flux profiles on divertor plates in the Large Helical Device

Masuzaki

Kobayashi

Morisaki

et al. 2009

Journal of Nuclear Materials

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Particle flux profiles on the divertor plates and the electron temperature profiles in the scrape-off layer (SOL) in the Large Helical Device (LHD) heliotron were investigated with EMC3-EIRENE code. These profiles are modified during a discharge due to the changes of the edge plasma density and temperature those can cause the change of transport coefficient.Comparison of the edge electron temperature profiles between the measurements and the simulations revealed that the cross field transport coefficients in the LHD scrape-off layer depend on plasma parameters, especially electron temperature. For the particle flux profile on the divertor plates, the absolute value of the simulation results with the transport coefficient consistent with the edge temperature profile analysis were well agree with the experimental data, though the profile shapes of experimental data were not necessarily reproduced well by the simulation.

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The divertor plasma characteristics in the Large Helical Device

Cited by 130 publications

References 28 publications

Poloidal Distribution Measurement of Inward Neutral Flux in LHD

Poloidal Distribution Measurement of Inward Neutral Flux in LHD

Thirty-Minute Plasma Sustainment by ICRF, EC and NBI Heating in the Large Helical Device

Investigation on the influence of plasma properties and SOL transport on the particle flux profiles on divertor plates in the Large Helical Device

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