Microstructure observations on butt joint composed of Nb3Sn CIC conductors

To precisely evaluate a butt joint technology for the JT-60SA CS coils, microstructure observations on the butt joint composed of Nb3Sn CIC conductors were conducted using a FE-SEM. As a sample for the observations, the butt joint sample utilized in the joint resistance measurement was used. During the sample fabrication, the butt joint sample was heated up to about 920 K from room temperature for diffusion bonding after heat treatment for Nb3Sn production. Then, the sample was subjected to the cycles of electromagnetic force in the joint measurement.The observation results indicated that Nb3Sn strands and a copper sheet were butted properly at the interface of the butt joint. In addition, there were hairline cracks in the Nb3Sn layers of the strands near the interface. To investigate a cause of the crack initiation, the stresses generated in the butt joint under same conditions were analyzed using a simple model. As a result, the cracks would occur with an axial compressive stress generated by the butt joint fabrication

Effects of titanium concentration on microstructure and mechanical properties of high-purity vanadium alloys

Effects of Ti concentration on microstructure and mechanical properties of high-purity V-4Cr-xTi alloys have been studied by means of scanning electron microscopy, transmission electron microscopy, Vickers hardness and tensile tests. Results show that precipitation occurs with 1 wt% Ti addition and above, whose diameter gradually increases as Ti concentration rises. Vickers hardness and tensile strength increase with increasing Ti concentration. Moreover, strengthening mechanisms consisting of solid solution strengthening (σSS), grain boundary strengthening (σGB), and precipitation strengthening (σP) are theoretically estimated. The strength contribution sequence is σSS > σGB > σP. Solid solution strengthening from Ti increases with increasing Ti concentration, and precipitation strengthening is not significantly dependent on Ti concentration. Additionally, 1 wt% Ti is probably sufficient to scavenge the interstitial impurities and provide comparable precipitation strengthening with V-4Cr-4Ti alloy

Validation of the plasma-wall interaction simulation code ERO2.0 by the analysis of tungsten migration in the open divertor region in the Large Helical Device

Tungsten migration in the open divertor region in the Large Helical Device is analyzed for validating the three-dimensional plasma-wall interaction simulation code ERO2.0. The ERO2.0 simulation reproduced the measurement of localized tungsten migration from a tungsten-coated divertor plate installed in the inboard side of the torus. The simulation also explained the measurement of the high tungsten areal density in the private side on a carbon divertor plate, next to the tungsten-coated divertor plate, by the tungsten prompt redeposition in plasma discharges for a low magnetic field strength in a counterclockwise toroidal direction. However, the simulation disagreed with the measurement of low tungsten areal density on the plasma-wetted areas on the carbon divertor plates, which indicated that the actual erosion rate of the redeposited tungsten should be much higher than that used in the ERO2.0 code

Investigation of remaining tritium in the LHD vacuum vessel after the first deuterium experimental campaign

Remaining tritium in the vacuum vessel after the first deuterium plasma experimental campaign conducted over four months was investigated in the large helical device (LHD) for the first time in stellarator/heliotron devices by using the tritium imaging plate technique. In-vessel components such as divertor tiles and first wall panels, and long-term material probes retrieved from the vacuum vessel were analyzed. The in-vessel component in which tritium remained most densely is the baffle part of divertor tiles made of graphite retrieved from the inboard-side divertor. Asymmetric tritium retention is observed on divertor tiles located at magnetically symmetric positions, and can be attributed to the toroidal field direction dependence of the asymmetric loss of energetic tritons generated by deuterium–deuterium nuclear fusion reactions. On the first wall, tritium remained in a deposited layer, which mainly consists of carbon

Studies of dust transport in long pulse plasma discharges in the large helical device

Three-dimensional trajectories of incandescent dust particles in plasmas were observed with stereoscopic fast framing cameras in a large helical device. It proved that the dust is located in the peripheral plasma and most of the dust moves along the magnetic field lines with acceleration in the direction that corresponds to the plasma flow. ICRF heated long pulse plasma discharges were terminated with the release of large amounts of dust from a closed divertor region. After the experimental campaign, the traces of exfoliation of carbon rich mixed-material deposition layers were found in the divertor region. Transport of carbon dust is investigated using a modified dust transport simulation code, which can explain the observed dust trajectories. It also shows that controlling the radius of the dust particles to less than 1 mm is necessary to prevent the plasma termination by penetration of dust for the long pulse discharges. Dust transport simulation including heavy metal dust particles demonstrates that high heating power operation is effective for shielding the main plasma from dust penetration by an enhanced plasma flow effect and a high heat load onto the dust particles in the peripheral plasma. It shows a more powerful penetration characteristic of tungsten dust particles compared to that of carbon and iron dust particles

Characterization of Ion Cyclotron Wall Conditioning Using Material Probes in LHD

The ion cyclotron wall conditioning (ICWC) is one of the conditioning methods to reduce impurities and to remove tritium from the plasma facing components. Among the advantages of ICWC are the possible operation under strong magnetic field for fully torus area based on the charge exchange damage observed in thin SS samples arranged on a hexahxedron block holder with three different facings, the areas influenced by ICWC is estimated. On the plasma facing area of the material holder, high density of helium bubbles is observed by transmission electron microscope (TEM). But the other areas show no observable damage. The fact that the bubble were observed only in a sample facing the plasma implies that the effective particles, most probably charge exchange neutrals come to the wall straightly Thus, cleaning of the surfaces un-exposed to plasma directly and those in shadow area is difficult by ICWC

Current Status of Large Helical Device and Its Prospect for Deuterium Experiment

Achievement of reactor relevant plasma condition in Helical type magnetic devices and exploration in its related plasma physics and fusion engineering are the aim of the Large Helical Device (LHD) project. In the recent experiments on LHD, we have achieved ion-temperature of 8.1 keV at 1 × 1019 m−3 by the optimization of wall conditioning using long pulse discharge by Ion Cyclotron Heating (ICH). The electron temperature of 10 keV at 1.6 × 1019 m−3 was also achieved by the optimization of Electron Cyclotron Heating (ECH). For further improvement in plasma performance, the upgrade of the Large Helical Device (LHD), including the deuterium experiment, is planned. In this paper, the recent achievements on LHD and the upgrade of LHD are described

Stable sustainment of plasmas with electron internal transport barrier by ECH in the LHD

The long pulse experiments in the Large Helical Device has made progress in sustainment of improved confinement states. It was found that steady-state sustainment of the plasmas with improved confinement at the core region, that is, electron internal transport barrier (e-ITB), was achieved with no significant difficulty. Sustainment of a plasma having e-ITB with the line average electron density ne_ave of 1.1 × 1019 m−3 and the central electron temperature Te0 of ∼3.5 keV for longer than 5 min only with 340 kW ECH power was successfully demonstrated

Progress of long pulse discharges by ECH in LHD

Using ion cyclotron heating and electron cyclotron heating (ECH), or solo ECH, trials of steady state plasma sustainment have been conducted in the superconducting helical/stellarator, large helical device (LHD) (Ida K et al 2015 Nucl. Fusion 55 104018). In recent years, the ECH system has been upgraded by applying newly developed 77 and 154 GHz gyrotrons. A new gas fueling system applied to the steady state operations in the LHD realized precise feedback control of the line average electron density even when the wall condition varied during long pulse discharges. Owing to these improvements in the ECH and the gas fueling systems, a stable 39 min discharge with a line average electron density ne_ave of 1.1  ×  1019 m−3, a central electron temperature Te0 of over 2.5 keV, and a central ion temperature Ti0 of 1.0 keV was successfully performed with ~350 kW EC-waves. The parameters are much improved from the previous 65 min discharge with ne_ave of 0.15  ×  1019 m−3 and Te0 of 1.7 keV, and the 30 min discharge with ne_ave of 0.7  ×  1019 m−3 and Te0 of 1.7 keV