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To identify the cause for an enhanced D retention in RAFM (reduced-activation ferritic/martensitic) steels with decreasing incident D fluence, ϕD, CLAM steel samples were exposed to D plasma at a low ϕD ∼ 3 × 1023 m−2 in the PISCES-A linear plasma device. At this low ϕD, a Cr-rich surface layer with a thickness of ∼ 10 nm still remains, confirmed with cross-sectional EDX (energy dispersive x-ray spectroscopy) elemental mapping. TEM (transmission electron microscopy) observations of a plasma-exposed surface reveal the formation of D bubbles inside the Cr-rich surface layer, while no D bubbles are seen in an unexposed surface. Furthermore, depth profiling using GDOES (glow discharge optical emission spectroscopy) detects D in the Cr-rich surface layer. These results demonstrate that the D retention at low ϕD is enhanced by D atoms trapped in bubbles created inside the Cr-rich surface layer.

In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

The transition of a detached to attached plasma experiment has been performed in GAMMA 10/PDX by applying an additional plasma heating pulse of electron cyclotron heating (ECH). In a plasma detachment experiment, a short pulse (25 ms) of ECH was applied at the east plug-cell to examine the effects of electron heating on the plasma parameters in the D-module. It was found that the ion flux increased significantly during ECH injection. In the ECH injection period, the ion flux increased with the increasing impurity injection, which indicates the impurity ion components are enhanced by ionization in the D-module due to the application of ECH. During ECH injection period, the increase of the ion flux near the corner of the target plate shows a clear dependence on the gas species. The 2D visible emission has been captured by the high-speed camera and a bright emission near the corner of the target has been observed. The spectroscopic measurement in the D-module shows that the impurity ion emission increases remarkably during ECH injection. These results indicate the detached plasma change to the attached state. Keywords: GAMMA 10/PDX, Divertor simulation experiment, Plasma detachment, Detached to attached transition

Three sets of tungsten coated graphite divertor tiles (VPS-W tiles) were installed in the closed helical divertor of the Large Helical Device (LHD) instead of the graphite divertor tiles in the 2012FY plasma campaign for examining the plasma wall interaction (PWI) with divertor plasma. The first wall panels and divertor tiles of the LHD consist of stainless steels (SUS316L) and graphite, respectively. The carbon based mixed-material deposition layer including a very small amount of Fe element has been formed on not only the divertor tiles but also the first walls near the divertor tiles through the PWI processes. Such a mixed layer often causes undesirable influences for maintaining a long pulse discharge in LHD, e.g., changing the particle recycling properties and dust generation. After the single plasma campaign of the 2012FY, we confirmed drastic suppression of the mixed-material deposition layer on the first wall panels just under the VPS-W tiles. On the other hand, carbon based mixed-material deposition layer was formed on the VPS-W tile surface, where the amount of the hydrogen retention was estimated to be over ∼4 × 1021 H/m2. If we would want further suppression of the hydrogen retention on the VPS-W tiles, the plasma facing components should be replaced to a full metal wall to avoid formation of the carbon co-deposition layer. Keywords: Tungsten divertor, Dust, Hydrogen retention

Long pulse plasma discharges in the Large Helical Device have been often interrupted by iron dust emission induced by electric arcing on the surface of the vacuum vessel. The iron ions in the peripheral plasma induced by the dust emission enhance physical sputtering and self-sputtering on the divertor plates, which can interrupt the long pulse plasma discharges by radiation collapse. The impurity transport simulation for a tungsten divertor configuration is performed using a three-dimensional edge plasma simulation code (EMC3-EIRENE) under the condition where the iron ions produced by the dust emission cause the sputtering on the tungsten divertor plates. The simulation shows that the sputtered tungsten significantly increases the radiation power by a factor of about 20 compared to that for the carbon divertor configuration in a high plasma temperature condition. The simulation reveals that plasma discharge operation with a high plasma density is desirable for the tungsten divertor. In this operational regime, the radiation power by tungsten ions is significantly reduced by the combined effect of the suppression of the sputtered tungsten and the control of the tungsten ion accumulation by the reduced thermal force in the peripheral plasma. Keywords: Tungsten, EMC3-EIRENE, Long pulse discharge, Sputtering, Dust

In this study, global particle balance of helium (He) long pulse discharge with graphite divertor in the Large Helical Device (LHD) and He retention in graphite from lab-scale experiments were compared in order to determine the role of the graphite divertor in He retention on the LHD wall. Global He particle balance analysis was conducted in long-pulse discharges in LHD with only turbo molecular pump. The analysis showed that static He retention was ∼2.9 × 1022 He and the ratio of retained He (= retention over fluence) was ∼0.45%. In the lab-scale study, He retention was measured by Thermal Desorption Spectroscopy (TDS) after ion irradiation or plasma exposure. The ratio of retained He were 0.1 ∼ 0.5% in all conditions, which was well consistent with LHD results. Therefore, it may be concluded that graphite divertor tiles have an important role in He absorption in the LHD wall at the initial phase of discharges.

In the present study, an additional deuterium (D) ion irradiation was performed against long-term samples mounted on the helical coil can and in the outer private region in the LHD during the 17th experimental campaign. Based on the release behavior of the D and hydrogen (H) retained during the experimental campaign, the difference of release behavior at the top surface and in bulk of modified surfaces is discussed. Almost all samples on the helical coil can were erosion-dominant and some samples were covered with boron or carbon, while a very thick carbon films were formed in the outer private region. In the erosion-dominant area, the D desorbed at much lower temperatures compared to that of H retained during the LHD plasma operation. For the samples covered with boron, the D tended to desorb at lower temperatures compared to H. For the carbon deposition samples, the D desorbed at much higher temperatures compared to no deposition and boron-covered samples, which was very similar to that of H. The D retention capabilities at the top surface of carbon and boron films were 2–3 times higher than no deposition area. The results indicate that the retention and release behavior at the top surface of the modified layer can be different from that of bulk substrate material. Keywords: Hydrogen retention, Desorption, Redeposition, Long-term sample, LHD

In order to confirm the formation of a tungsten fiberform nanostructure (W-fuzz) by helium plasma exposure in the large-sized plasma confinement device, the bulk tungsten with the size of 80 × 20 × 1.5 mm3 was inserted into the divertor leg position in the Large Helical Device (LHD). Then, it was exposed to the divertor plasma during the ultra-long pulse helium discharges with 10,190s in total. The width of the divertor plasma, incident ion energy and total fluence were expected to be ∼2cm, 100–200eV and ∼5 ×1025 He/m2 (strike position), respectively. The surface temperature of the tungsten specimen was monitored by IR camera. The typical surface temperature of the divertor strike point was estimated to be around 1900∼2300K. After the exposure, an initial growth phase of tungsten fiberform nanostructure (W-fuzz) was able to be identified on the tungsten surface. The finest initial growth phase of the W-fuzz structure was able to be identified on the central region of the divertor strike point, where retention amount of helium was estimated to be ∼8 ×1021 He/m2. This study is the first simultaneous evaluation of the W-fuzz growth and quantification of the helium retention in the large-sized plasma confinement device. Keywords: W-fuzz, TEM observation, Ion beam analysis, LHD

A new liquid metal divertor system named the REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed for the helical fusion reactor FFHR-d1. The REVOLVER-D is composed of molten tin shower jets stabilized by internal flow resistances of wire/tape/chain. These shower jets are inserted into the ergodic layer surrounding the main plasma. Tin is selected as the liquid metal because of its low melting point, low vapor pressure, low material cost, and high safety. The liquid metal pumps, cryopumps, and turbo molecular pumps are installed in the central vacuum vessel connected to the main vacuum vessel via 10 inner ports equipped with maze neutron shields. Central solenoid coils made of high-temperature superconductors are installed inside the central vacuum vessel to shield the pumps from the strong magnetic field. The REVOLVER-D has a good possibility to satisfy important characteristics required for the divertor system in a fusion reactor, that is, high heat load tolerance, high maintainability, sufficient vacuum pump speed, high level of safety, and a small amount of radioactive wastes.

In this study, the effect of ELM-like thermal shocks on He induced surface morphology were investigated. A W sample was exposed to pure He plasma. He ion incident energy, flux and fluence were 80 eV, 1 × 1022 /m2/s and 3 × 1025 /m2, respectively. Irradiation temperature were approximately 470 and 1100 K. Then, thermal shocks were applied on the sample using focused electron beam. The peak heat flux, pulse duration and base temperature were 500 MW/m2, 500 μs and R.T., respectively. After the thermal shocks, He induced morphologies such as holes and periodic undulations were flatten completely at the region exposed to the highest heat flux. At the peripheral regions of the electron beam spot, the hole density increase or partial flattening of morphologies were observed. These results suggested that in order to anticipate surface morphology with He irradiation and ELMs, the peak temperature should play a more important role than base temperature. Keywords: Tungsten, Divertor, He induced surface morphology, Thermal shocks

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

The subcooled boiling heat transfer and the steady-state critical heat flux (CHF) in a vertical circular tube for the liquid Reynolds numbers ( Re d = 3.65 × 10 4 –3.08 × 10 5 ) and the flow velocities ( u = 3.95–30.80 m/s) were systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter ( d = 6 mm) and heated length ( L = 59.5 mm) was used in this work. The boiling initiation noise of outer surface of the test tube in the open air was simultaneously measured up to CHF point by the sound level meter (SLM) and the microphone of a video camera (MP). The outer surface temperatures of the SUS304 test tube with heating were also observed by an infrared thermal imaging camera (ITIC) and the color temperatures of outer surface of the test tube in the open air were observed by a video camera (VC). The subcooled boiling heat transfer and CHF for SUS304 circular tube were compared with the values calculated by authors’ and other researchers’ correlations for the subcooled flow boiling heat transfer. The influences of flow velocity on the boiling initiation surface heat flux, the boiling initiation surface superheat, the subcooled boiling heat transfer and the CHF were investigated into details based on the experimental data. At the flow velocities higher than 13.3 m/s, boiling initiation surface heat fluxes were close to the CHFs and surface superheats at the CHF were over to the homogeneous spontaneous nucleation temperature as well as the lower limit of the heterogeneous spontaneous nucleation temperature. The dominant mechanism of the subcooled water flow boiling CHF on the SUS304 circular tube was discussed at high liquid Reynolds number.

The subcooled boiling heat transfer and the steady state critical heat flux (CHF) in a vertical circular tube for the liquid Reynolds numbers (Red = 2.77 × 104–3.08 × 105) and the flow velocities (u = 3.95–30.80 m/s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d = 6 mm) and heated length (L = 59.5 mm) is used in this work. The outer surface temperatures of the SUS304 test tube with heating are observed by an infrared thermal imaging camera and a video camera. The subcooled boiling heat transfers for SUS304 test tube are compared with the values calculated from correlations due to other researchers for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated in detail based on the experimental data. Nucleate boiling surface superheats at the CHF are close to the lower limit of the heterogeneous spontaneous nucleation temperature and the homogeneous spontaneous nucleation temperature. A suggestion as to what the dominant mechanism is for the subcooled flow boiling CHF on the SUS304 circular tube is made at high liquid Reynolds number. On the other hand, the RANS equations (Reynolds Averaged Navier–Stokes Simulation) with k–ε turbulent model in a circular tube of a 3 mm in diameter and a 526 mm long are numerically solved for heating of water on heated section of a 3 mm in diameter and a 67 mm long with various thicknesses of conductive sub-layer by using PHOENICS code under the same conditions as the experimental ones previously obtained and with temperature dependent thermo-physical fluid properties. The Platinum (Pt) test tube of inner diameter (d = 3 mm) and heated length (L = 66.5 mm) was used in this experiment. The thicknesses of conductive sub-layer from non-boiling regime to CHF are measured. The thicknesses of conductive sub-layer at the CHF point are predicted for various flow velocities. The experimental values of the CHF are also compared with the corresponding theoretical values of the liquid sub-layer dry-out models suggested by other researchers, respectively. A suggestion as to what the dominant mechanism is for the subcooled flow boiling CHF on the Pt circular tube is made at high liquid Reynolds number.

The three-dimensional trajectories of dusts have been observed with two stereoscopic fast framing cameras installed in upper and outer viewports in the Large Helical Device (LHD). It shows that the dust trajectories locate in divertor legs and an ergodic layer around the main plasma confinement region. While it is found that most of the dusts approximately move along the magnetic field lines with acceleration, there are some dusts which have sharply curved trajectories crossing over the magnetic field lines. A dust transport simulation code was modified to investigate the dust trajectories in fully three dimensional geometries such as LHD plasmas. It can explain the general trend of most of observed dust trajectories by the effect of the plasma flow in the peripheral plasma. However, the behavior of the some dusts with sharply curved trajectories is not consistent with the simulations.

The heat flux at the divertor in a fusion reactor is considered to have a peak of >10 MW/m 2 . In a design study of the helical reactor FFHR-d1, the feasibility of employing a copper alloy for divertor cooling pipes was investigated; however, radiation in the divertor area would quickly damage the copper alloy. The neutron load on the divertor can be reduced by a blanket arrangement; nevertheless, in the present divertor structure, irradiation damage of materials on the inboard side of the torus remains relatively high. If the divertor could be moved to an area receiving much less radiation, then the lifetimes of divertor materials should increase. In this paper, a novel divertor structure is introduced in which the coil-support structure is modified to create a region receiving relatively low amounts of radiation without changing the geometry of the helical or vertical field coils. Using this proposed design would increase the lifetime of the copper alloy in divertor components to more than an estimated six years. In addition, the divertor could be accessed from either the upper or lower sides of the device, simplifying maintenance.