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Steady-state sustainment of divertor detachment with multi-species impurity seeding in LHD
Get PDF0000-0003-1586-1084Multi-species impurity seeding is an advanced operation scenario to mitigate the divertor heat load for the realization of future fusion reactors. In the large helical device, divertor detachment is successfully sustained using higher-Z (krypton, Kr) and lower-Z (neon, Ne) superimposed seeding. Emission from Kr impurities is drastically enhanced if it is followed by Ne seeding. Plasma radiation can be enhanced even at the upstream region in the edge plasma compared with Ne only seeded plasmas with suppression of impurity accumulation toward the central plasma. The characteristics of divertor heat load reduction and energy confinement are comparable between the Kr + Ne seeding and Ne only seeding under the same radiation fraction. However, while the detachment in Ne only seeding is transient, the detachment in Kr + Ne seeding is stable. It indicates that multi-species impurity seeding can be competitive for steady-state operation although further investigation is desired about the balance between divertor heat load reduction, impurity screening, and confinement degradation. The Kr emission enhancement is strongly affected by electron density and temperature at the last closed flux surface resulting in impurity penetration.journal articl
Removal of tritium from vacuum vessel by RF heated plasmas in LHD
Get PDF0000-0001-9941-1958The tritium removal from the vacuum vessel is one of the key issues for the realization of a fusion reactor. As in situ tritium removal techniques from the plasma facing materials, helium glow discharge wall conditioning (GDWC) and RF heated plasma due to electron cyclotron wall conditioning (ECWC) and ion cyclotron wall conditioning (ICWC) are applied with Large Helical Device (LHD) which remains tritium produced by deuterium plasma experiment. As the result, it suggests that ICWC could remove the tritium more efficiently than the ECWC and the GDWC. In the ICWC operation, the high energy particles induced by the minority ion heating mode as the standard ICRF heating operation may play an important role in tritium removal from the plasma-facing materials due to the particle's bombardment.journal articl
Engineering study on lithium isotope separation by ion exchange chromatography
Get PDF0000-0002-1437-2273We have been studying the lithium isotope separation by ion exchange chromatography for the engineering development. We performed the lithium isotope separation experiments by chromatography. From the results, we estimated the migration lengths required to obtain the 20 %, 50 %, and 90 % enriched lithium-6, the required lengths are 340, 1600, and 5800 m, respectively. We studied the detection methods of lithium adsorption band boundary in order to design the circular type chromatographic equipment, which type equipment is required for the industrial plant. Based on these investigations, we designed the automatic circular type chromatographic equipment with the boundary interface detector and manufactured the prototype equipment.journal articl
Study on Assembly of TF Coil and Vacuum Vessel for Fusion DEMO
Get PDFITER is in the integration phase and a fusion DEMO is being designed conceptually. The size of the DEMO will be around 1.5 time larger than the ITER and the weight will be about three times larger than the ITER. The TF coil and the vacuum vessel of the DEMO will become very heavy and extremely large. Although the central support structure will be used in the ITER for the assembly of TF coils and vacuum vessels, the central cylinder concept is proposed in this study to avoid the huge concentrated load at the center of the device floor during assembly. In addition, the welding joint between the vacuum vessels is discussed under the limited conditions supposed on the DEMO. Since the vacuum vessel will become the nuclear boundary, all weld lines will be expected to be examined by ultrasonic testing and/or radiographic testing to ensure the soundness of the welds.journal articl
Influence of Cu-Y Compound Content on the Microstructure of Cu-Y2O3 Dispersion Strengthened Alloys Synthesized by MA and HIP Process
Get PDF0000-0003-1397-9556Oxide Dispersion Strengthened Cu alloy (ODS-Cu) with Y2O3 has great application potential in the field of fusion. In the previous fabrication method by adding metal Y as the source of Y2O3, severe sticking occurred during mechanical alloying (MA) and huge Y particles still remained after MA. Considering that Cu-Y compounds are more brittle than pure Y, which is expected to resolve the sticking issue, and have lower Y enrichment, which will make it easier to form uniform Y distribution, ODS-Cu with various content of Y by adding Cu2Y or Cu6Y were fabricated through MA and Hot isostatic pressing (HIP) process. A comparative analysis was made for the samples with addition of Cu2Y and Cu6Y. The results showed that, compared with Cu2Y, Cu6Y is easier to form uniform Y distribution. The most likely reason is that the Cu6Y is more brittle and less Y abundance than Cu2Y. The sample with 0.39 wt% Y with Cu6Y addition has the highest Vickers hardness, possibly because of solution strengthening caused by higher content of interstitial O, and better electrical conductivity than the sample with 1.19 wt% Y with Cu6Y possibly because of the more uniform formation of Y2O3 and the absence of precipitation phases.journal articl
Plasma irradiation experiment on the metal pebble flow in the TPDsheet-U
Get PDFA new divertor concept REVOLVER-D2 that uses metal pebbles as a divertor target has been proposed for the LHD-type helical fusion reactor FFHR-c1. Demonstration of sufficient plasma shielding rate is one of the important issues to show the feasibility of the REVOLVER-D2. For this purpose, an experiment on the plasma irradiation to the metal pebble flow has been conducted in the sheet plasma device TPDsheet-U in Tokai University. Reduction of the ion flux and electron density in the metal pebble flow was observed and it was confirmed that the metal pebble flow can play the role of the divertor target. The measured shielding rate is consistent with the prediction by the model based on the Lambert-Beer-Bouger’s law. Achievement of the shielding rate of > 72 % is expected in the FFHR-c1 using the nozzle with the size of > 310 mm × 310 mm and the pebble flow with the diameter of 5 mm.journal articl
Simultaneous Observation of Silicon and Boron Impurity Behaviors in the Core Region of a Mid-Density LHD Plasma
Get PDF0000-0003-1682-1519Line emissions from both silicon (Si) and boron (B) impurity ions introduced by a single tracer-encapsulated solid pellet (TESPEL) containing silicon hexaboride (SiB6) powders were successfully observed using the extreme ultraviolet (EUV) spectrometer and charge-exchange spectroscopy (CXS) technique in the Large Helical Device. The CXS diagnostic shows clearly that a hollow radial profile of fully ionized B impurities was created immediately after the TESPEL injection, and such a hollow profile was relaxed with time. At the same time, Li-like emissions from the highly ionized Si impurities were also observed with the EUV spectrometer, SOXMOS. Therefore, the decay times of these impurities could be estimated under the same plasma conditions. The estimated decay time of the Si impurities, τSi = 0.12 ± 0.01 s, was found to be slightly longer than that of the B impurities, τB = 0.09 ± 0.01 sjournal articl
Mode Structure of Locked-Mode-Like Instability in LHD and Its Effects on Confinement Degradation
Get PDF0000-0003-3754-897XThe internal mode structure of a precursor with the tearing-parity structure of the locked-mode-like instability is investigated. For the first time, the Thomson scattering system with high temporal and spatial resolution enables us to find the non-rotating temperature flattening region in the torus outboard side in addition to the rotating island during the slowing-down phase. Additionally, the width of the flattening region is 10% normalized by the plasma minor radius. Furthermore, the radial profiles of the pressure degradation during the slowing-down phase are evaluated. At the beginning of the slowing-down phase, the pressure degradation area is located slightly inside the resonant surface. After that, the peak location moves to the core region. Finally, the degradation area peaks at the plasma center.journal articl
Application of Divertor Pumping to Long-Pulse Discharge for Particle Control in LHD
Get PDF0000-0001-5522-3082Divertor pumping was applied to plasma discharges for superior fuel particle control in the Large Helical Device (LHD). The LHD is equipped with two different pumping systems. One is the main pumping system, in which the pumping speed is 260 m3/s in hydrogen. The other pumping system is the divertor pumping system in which the pumping speed is 70 m3/s in hydrogen. Divertor pumping was applied to 40-second long pulse Electron Cyclotron Heating (ECH) discharges to assess the improvement in particle control provided by divertor pumping. The results show that without divertor pumping, the electron density was not controlled by gas puffing using the feedback signal of line-averaged electron density. Then, the plasma confinement deteriorated, finally leading to radiation collapse. On the other hand, with divertor pumping, the density was well-controlled by gas puffing using the feedback signal. The results indicate that divertor pumping is one of the key tools for controlling the particles in fusion plasmas.journal articl
Measurement of Decay Time Constant of Shielding Current in ITER-TF Joint Samples
Get PDFJoint sample tests have been carried out as a qualification test for ITER Toroidal Field (TF) coils. The joint sample comprises two short TF conductors that have "twin-box" joint terminals at both ends. The lower joint is a testing part that is a full size joint of the TF coils. Hall probes are attached on the lower joint box at around the center of the external field coil of the test facility. The magnetic field induced by shielding currents in the joint can be estimated from the difference between the measured magnetic field strength and the magnetic field generated by the external field coil. The magnetic field by the shielding currents during shut-off of the external field coil from -1.0 T is evaluated for six samples. The decay time constants of the shielding currents are gradually elongated with decrease of the shielding currents in all the samples. In comparison with simulation results, it is considered that the main shielding current flows in superconducting cables in the two conductors with crossing the jointed plane and that the joint resistance is decreased at low total current.journal articl