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A predictor model of radiative collapse of stellarator-heliotron plasmas has been developed by means of a machine learning technique and the feature of radiative collapse has been extracted with sparse modeling. The dataset used for training the model is constructed based on density ramp-up experiments in the Large Helical Device. As a result of feature extraction, the line averaged electron density, visible line emissions of CIV and OV, and the electron temperature at the edge have been selected as key parameters of radiative collapse. The likelihood of occurrence of radiative collapse has been quantified by using these parameters and this likelihood has been assessed in terms of predicting capability of the occurrence of radiative collapse. The collapse likelihood also implies the underlying physics of radiative collapse, therefore, the knowledge obtained by this data-driven study is expected to facilitate elucidation of the physics of the radiative collapse. In validation with discharges outside of the dataset, the predictor based on the likelihood has predicted over 85% of radiative collapse about 100 ms prior to this event on average while about 5% of stable discharges have been detected falsely as collapse discharges. The discharges in which the predictor made faults are discussed in order to investigate the cause of failure.

A laboratory-scale inductively coupled plasma apparatus investigated the behavior of liquid Sn–Bi–Li–Er alloy (SBLE) under hydrogen (H2) plasma exposure. By exposing the liquid SBLE to H2 plasma, the bubbling of liquid SBLE and pulsive H2 pressure increase was observed. Moreover, plasma density, i.e. ion flux, increased both bubble size and H2 pressure spike. The simultaneous measurement of a high-speed camera and mass spectrometry concluded that the bubble was formed by H2 accumulation, and ∼38% of hydrogen ion flux contributed to the H2 accumulation. Lithium hydride may be considered the bubble sphere’s origin from X-ray photoelectron spectroscopy measurement of droplets produced from the ruptured bubble sphere. From the experimental results, the bubble formation mechanism was discussed.

In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu–0.3 wt%Al2O3) and tungsten (W) demonstrated superior fracture strength (∼200 MPa). This joint was fabricated by the direct brazing method between W and ODS-Cu using BNi-6 (Ni–11%P) filler material without any intermediate layer. This method was named as the improved or the advanced brazing technique. In the present study, deformation and fracture behavior of the joint after the three-point bending test was investigated. At first, it was found that the crack initiation points were dominantly in the W bulk, although it was not clear that the crack initiated from grain boundary or not. Secondly, the crack propagation proceeded mostly in the W bulk but tended to deflect towards the bonding layer. These results interpret the strength of the bonding interface is superior and the present bonding technique is applicable for severe environments such as a high heat flux divertor component on the fusion reactor. Based on the above physical and technological understanding, we successfully fabricated the large scale divertor mock-up which has twenty-eight plates of W with each size of 20 × 20 × 5 mm3.

A case of the application of Non Evaporable Getter (NEG) pumps to fusion devices is discussed. Compared with other pumping technologies used in fusion devices, the NEG pumps have advantages in terms of operation temperature at 100–200 °C (in which there is a high tolerance of the temperature increase due to a plasma radiation and/or an unexpected heat load from heating devices), low power consumption and compactness. Moreover, in the case of a power outage getters continue pumping and do not release hydrogen and its isotopes. The NEG pumps have been installed as an in-vessel pump into the divertor region in the Large Helical Device (LHD) for the first time. The pumping performance test shows that the effective pumping speed is 10 m3/s in hydrogen, which is close to the target value. In addition, the exposure of the NEG pumps to boronization and glow discharge treatments have been tested. While the boronization and glow discharge under noble gases do not show effect on the pumping performances of the NEG pumps, the establishment of an operational scenario including hydrogen and deuterium glow discharges consistent with other pumping facilities is still an aspect to be refined. However, the use of NEG pumps in LHD divertor is a milestone for the possibility of using this technology in future fusion devices.

Divertor particle flux signals on divertor plates were analyzed with the multivariable analysis technique, namely, proper orthogonal decomposition (POD), to characterize the indefinite divertor footprint in the Large Helical Device. To verify whether isotope effects exist in relation to the divertor plasmas, the POD outputs in hydrogen (H) and deuterium (D) plasmas were first compared by analyzing a number of discharges. It was found that the dominant components of the divertor particle flux profiles in H and D plasmas were similar, indicating no isotope effect. In addition, total reconstructed divertor particle fluxes in H plasmas were higher than those in D plasmas, suggesting an isotope effect. Furthermore, the footprint-profile modification with from the SOL-side to the private-side peak due to the edge electron pressure gradient in H and D plasmas were slightly different, relating to an isotope effect. Keywords: Divertor footprint, Isotope effect, Langmuir probe, Proper orthogonal decomposition, LHD

We have measured extreme ultraviolet (EUV) and soft X-ray emission spectra of krypton (Kr) ions in the Large Helical Device (LHD) plasmas with Kr gas seeding intended for mitigating divertor heat loads. Several isolated lines or quasi-continuum bands from ion stages in the range of Kr5+–Kr25+ have been identified in a variety of wavelength ranges. The time trends of the line intensities indicate major contributions of ions with M-shell (n = 3) outermost electrons to the enhanced total radiation in the region closer to the core plasma. A line intensity ratio analysis for Kr25+ implies an unignorable contribution of the recombining component for the emissivity from this ion stage. A difference between the trends of neon (Ne) and Kr is also observed in a discharge with a combination of Kr and Ne seedings intended for more effective radiation enhancement. Keywords: Collisional-radiative model, Radiation enhancement, Impurity seeding, Large Helical Device, Spectroscopy, Krypton

An asymmetric plasma particle load on divertor tiles located at positions of symmetric magnetic field lines structure has been investigated in the Large Helical Device (LHD). The asymmetry reverses with reversing the toroidal magnetic field direction. This suggests that drifts affect the asymmetry. It is found that the degree of the asymmetry depends on the electron temperature in the edge plasma and on the toroidal magnetic field strength. The degree of the asymmetry becomes large and then saturates with an increase of electron temperature. In the lower magnetic field discharge, the degree of the asymmetry is larger. The mass dependence of the asymmetry is not clear among hydrogen, deuterium and helium. Results presented suggests that E × B drift plays a role in formation of the asymmetry. Keyword: LHD, Divertor, Asymmetric particle load, Drift

The core plasma confinement during detachment phase is investigated in the discharges with application of resonant magnetic perturbation (RMP) field in LHD. The RMP application creates a remnant magnetic island in the edge stochastic layer, which largely changes the plasma parameter profiles including impurity radiation. The electron temperature and pressure profiles are flattened at the island, while the electron density is slightly peaked at the edge of the island. The estimated impurity radiation profile is enhanced and fixed around the magnetic island during the detached phase, where the discharge is stably sustained with controlled level of radiation. Without RMP, the radiation penetrates the confinement region, leading to radiation collapse. It is found that in the case of the RMP application the plasma stored energy increases discontinuously at the detachment transition. In spite of the reduced effective plasma volume caused by the edge magnetic island and by the enhanced radiation there, the central plasma pressure finally exceeds the case without RMP. This is caused by the pressure profile peaking at the central region in the case with RMP. These results indicate clear change of core plasma confinement during the detached phase with RMP.

Tungsten coating with a thickness of 1 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W), which is a leading structural material candidate for DEMO, have been produced by Atmospheric Plasma Spraying (APS) and Vacuum Plasma Spraying (VPS). Heat loading experiments on thermal response and fatigue using an electron beam have been carried out on W-coated F82H brazed on oxygen free high purity copper (OFHC) block with a cooling tube to evaluate their possibility as a plasma-facing armor in the fusion device. In addition, quantitative analyses about temperature profiles and thermal stress have been carried out using FEA. Thermal response experiments show that the temperatures increased monotonically with increasing heat flux. Surface temperature of the VPS-W/F82H/OFHC is always lower than that of the APS-W/F82H/OFHC. Surface modification, exfoliation and crack are not formed by thermal fatigue experiments up to 200 cycles at a heat flux of 3.2 MW/m2. Thermal response experiments under steady state condition have been successfully modeled by FEA. Interfacial strength of VPS-W and F82H is also discussed using the FEA results.

Long pulse discharges in the Large Helical Device have often been interrupted by large amounts of dust particle emission from the divertor region caused by the exfoliation of carbon-rich mixed material deposition layers. The plasma wall interaction code ERO2.0 has provided the simulation results of the three-dimensional distribution of the carbon flux density in the divertor region which is quite reasonable with the observed distribution of the carbon-rich deposition layers. The code has also succeeded in reproducing the reduction of the carbon deposition layers on dome plates by changing the target plate configuration in the divertor region. The ERO2.0 simulations have also successfully explained dust particle emission from the inboard side near the equatorial plane for the new target plate configuration at the termination of a long pulse discharge. These simulation results prove that the ERO2.0 code is applicable to predicting the possible position from where the dust particles are released, and to designing an optimized divertor configuration for performing stable long pulse discharges with controlled dust particle emission.

In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed.

The transition between isotope-mixing and nonmixing states in hydrogen-deuterium mixture plasmas is observed in the isotope (hydrogen and deuterium) mixture plasma in the Large Helical Device. In the nonmixing state, the isotope density ratio profile is nonuniform when the beam fueling isotope species differs from the recycling isotope species and the profile varies significantly depending on the ratio of the recycling isotope species, although the electron density profile shape is unchanged. The fast transition from nonmixing state to isotope-mixing state (nearly uniform profile of isotope ion density ratio) is observed associated with the change of electron density profile from peaked to hollow profile by the pellet injection near the plasma periphery. The transition from nonmixing to isotope-mixing state strongly correlates with the increase of turbulence measurements and the transition of turbulence state from TEM to ion temperature gradient is predicted by gyrokinetic simulation.

The transport of impurities supplied by a multi-species impurity powder dropper (IPD) in the large helical device (LHD) is investigated using a three-dimensional peripheral plasma fluid code (EMC3-EIRENE) coupled with a dust transport simulation code (DUSTT). The trajectories of impurity powder particles (Boron, Carbon, Iron, and Tungsten) dropped from the IPD and the impurity transport in the peripheral plasma are studied in a full-torus geometry. The simulation reveals an appropriate size of the impurity powder particles and an optimum operational range of the dust drop rates for investigating the impurity transport without inducing radiation collapse. The simulation also predicts a favourable plasma discharge condition for wall conditioning (boronization) using the IPD in order to deposit boron to high plasma flux and neutral particle density areas in the divertor region in the inboard side of the torus.

The turbulent heat transfer, the subcooled boiling heat transfer and the steady state CHF for a Pt-circular test tube of a 3 mm inner diameter and a 100 mm heated length are measured with a wide range of inlet subcooling and flow velocity at high liquid Reynolds number, i.e. Re-d = 3.01x10(4) to 1.43x10(5). The inner surface temperature of the Pt-circular test tube calculated by the steady one-dimensional heat conduction equation is compared with the values derived from authors' turbulent heat transfer correlation and with the numerical solutions of the RANS equations (Reynolds Averaged Navier-Stokes Simulation) of k-ε turbulence model for the flow velocities ranging from 4 to 21 m/s. The thicknesses of conductive sub-layer from non-boiling regime to CHF are measured by numerically analyzing the heat transfers with conductive sub-layer on forced convection and with thinner one dissipated by the evaporation on nucleate boiling. The thicknesses of viscous sub-layer on forced convection are estimated from the thicknesses of the conductive sub-layer and Prandtl numbers of the surface temperature on the heated surface. Furthermore, the thicknesses of conductive sub-layer at the CHF point are extrapolated from the measured values at various flow velocities. The experimental values of the CHF are also compared with authors' widely and precisely predictable correlations of critical heat flux during flow boiling of subcooled water and the corresponding theoretical values of the liquid sub-layer dry-out models suggested by other researchers, respectively. The authors' correlations and other researchers' theoretical values can represent the subcooled boiling CHFs obtained in this study within the ranges of -13.27 to 6.76% difference and - 32.51 to 13.16% one, respectively. A suggestion based on the experimental data as to what the dominant mechanism is for critical heat flux during flow boiling of subcooled water on a vertical circular tube is confirmed again at high liquid Reynolds number. The transitions to film boiling at the subcooled water flow boiling on the Pt test tube of d = 3 mm and L = 100 mm would occur due to the liquid sub-layer dry-out model at the steady-state CHF as well as those on the Pt test tube of d = 3 mm and L = 66.5 mm, but not due to the heterogeneous spontaneous nucleation and the hydro-dynamic instability.

Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W. Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy

The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.

Various incipient boiling phenomena for subcooled water flowing in a uniformly heated narrow tube were observed experimentally. The boiling signal was analyzed using the wavelet decomposition method. The boiling incipience of subcooled water in the narrow tube was recorded by a sound level meter at various flow velocities. A platinum tube was used as the experimental tube with an inner diameter of 1.0 mm. The length of the experimental tube was 23.2 mm. The tube was heated by the Joule effect using a direct current. The inlet temperature and flow velocities ranged 285–346 K and 2.5–14 m/s, respectively. The surface superheat ascended with an increase of the heat flux until the incipient boiling point was reached. The initial temperature overshoot did not appear as the outlet pressure increased. Since the existing correlations underestimated the incipient heat flux, a semi-empirical correlation of the boiling incipience was obtained based on the experimental data. The predicted value of the new correlation is in agreement with the experimental data within ±30%.

Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.

The reliable bonding between tungsten (W) and reduced activation ferritic/martensitic (RAFM) steel (JLF-1) was obtained by brazing with BNi-6 (Ni-11%P) filler material. In this work, a pure copper (Cu) interlayer was selected for the absorber of the residual stress, in which BNi-6 filler materials are inserted between W/Cu and Cu/JLF-1 interface. The stacked structure of the W/BNi-6/Cu/BNi-6/JLF-1 was subjected to the heat treatment procedure at 960℃. After the heat treatment procedure, fine joint structure without any crack and large size pore can be confirmed. In addition, several kinds of detailed analysis such as hardness testing and element distribution in the W/BNi-6/Cu/BNi-6/JLF-1 bonding structure, were performed. It was found that the Ni and P elements were preferentially located near each joint interface, and those elements caused a hardening effect of the bonding layer through some segregation mechanism. Also, the joint sample is subjected to the heat loading test by the electron beam device (ACT2) for evaluating the heat conducting characteristics of the joint. According to the temperature measurement during the heat loading, those hardening layers seemed to not cause the negative effects for degrading the joint properties.

In this study, the hydrogen retention properties of solid tin exposed to DC glow plasma were investigated using thermal desorption spectroscopy. The measurements were performed by varying the plasma exposure time from 0 (no exposure) to 40 min. The resulting retained hydrogen from the ion fluence of the plasma was quantified as ~10−3. The retained hydrogen increased with the exposure time.

We evaluated the two dimensional (2D) distribution of the divertor heat flux in LHD. The Infrared (IR) thermography was performed to measure the surface temperature at a divertor plate. The 3D heat conduction equation was solved using the finite element method (FEM) by taking into account the practical divertor geometry including the heat sink behind the graphite tile and the cooling system. The FEM analysis successfully reconstructs heat load distribution from the measured temperature pattern. Significant difference was found between the temperature and the heat load patterns. The FEM analysis shows that the highest heat deposition is observed at the strike line with 5–10 MWm−2 of ∼ 10 mm width in NBI heated discharge. In addition to the strike line there is also found the lower heat deposition region of ∼ 1 MW m−2 with wide channel width ∼ 30 mm. The detailed heat transport analysis inside the divertor components shows that the heat transport process is different between the strike line and the other region due to the heat channel width and to the divertor component structure. The comparison between the heat flux obtained by the thermography and that by the Langmuir probes shows reasonable agreement, except that the peak value of the heat load is higher in the IR thermography than in the Langmuir probe. By relaxing the assumption that electron temperature, Te, equal to be ion temperature, Ti, in the sheath model for the Langmuir probe analysis, the agreement becomes better for the case with Ti > Te.

Toroidal distributions of divertor particle flux during neon (Ne) and nitrogen (N2) seeded discharges were investigated in the Large Helical Device (LHD). By using 14 toroidally distributed divertor probe arrays, which were positioned at radially inner side where the divertor flux concentrates in the inward-shifted magnetic axis configuration, it is found that Ne puffing leads to toroidally quasi-uniform reduction of divertor particle fluxes; whereas toroidally localized reductions were observed with N2 puffing. The toroidally asymmetric reduction pattern with N2 puffing is strongly related to the magnetic field structure around the N2 puffing port. Assuming that nitrogen particles do not recycle, EMC3-EIRENE simulation shows similar reduction pattern with the experiment around the N2 puffing port. Keywords: Impurity seeding, LHD, Toroidal divertor probe arrays, Neon, Nitrogen, EMC3-EIRENE

Surface morphology and hydrogen isotope retention of W specimen melted with vertical displacement event-like heat load and subsequent deuterium (D) plasma exposure were studied. Applied heat loads using electron beam without raster scanning were about 190 and 230 MW/m2 in heat flux and 0.08, 0.12 and 0.16s in duration. After the heat load application, specimens showed apparent melting spots with grain growth or dense micrometer scale convex structure. Cracks were observed only in the part with the convex structure. D retention in the melted part of specimens was not significantly larger than in the reference specimen despite large changes of surface characteristics. Keywords: Tungsten, Retention, Melting, Divertor, High heat flux, NRA

Two different plasma termination processes by dust emission were observed in long pulse discharges in the Large Helical Device. One is a plasma termination caused by large amounts of carbon dust released from a lower divertor region. The other is termination caused by stainless steel (iron) dust emission from the surface of a helical coil can. The effect of the dust emission on the sustainment of the long pulse discharges are investigated using a three-dimensional edge plasma transport code (EMC3-EIRENE) coupled with a dust transport code (DUSTT). The simulation shows that the plasma is more influenced by the iron dust emission from the helical coil can than by the carbon dust emission from the divertor region. The simulation revealed that the plasma flow in divertor legs is quite effective for preventing dust from terminating the long pulse discharges. Keywords: Dust, EMC3-EIRENE, DUSTT, Long pulse discharge

In the tandem mirror GAMMA 10/PDX, molecular activated recombination (MAR) leading to plasma detachment has been observed by additional hydrogen gas injection to the divertor simulation plasma (i.e. end loss plasma) which is exposed to the V-shaped target in the divertor simulation experimental module (D-module). The temperature near the corner of the V-shaped target decreased from ∼23 eV to ∼2 eV as the neutral pressure in the D-module increased. A clear density rollover was observed at ∼2 Pa. A position of the density maximum moves to upstream of the plasma with increase in the neutral pressure and the density near the corner of the target decreases to detach the plasma from the target. After the occurrence of the density rollover, the Balmer β intensity decreases as with the density but the Balmer α intensity continues to increase, indicating the dissociative attachment process in MAR is more dominant than the ion conversion process although the rate coefficient of the former process is lower than that of the latter one, which is calculated by using a collisional radiative model. This would be caused by the MAR process related to triatomic hydrogen molecules which significantly contributed to the detachment process.

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...

A new temperature controlled material probe was designed for the exposure of tungsten samples to helium plasma in the LHD. Samples were exposed to estimated fluences of ∼1023 m−2 and temperatures ranging from 65 to 600 °C. Transmission Electron Microscopy analysis allowed the study of the impact of He irradiation under high temperatures on tungsten micro structure for the first time in real-plasma exposure conditions. Both dislocation loops and bubbles appeared from low to medium temperatures and saw an impressive increase of size (factor 4 to 6) most probably by coalescence as the temperature reaches 600 °C, with 500 °C appearing as a threshold for bubble growth. Annealing of the samples up to 800 C highlighted the stability of the dislocation damages formed by helium irradiation at high surface temperature, as bubbles and dislocation loops seem to conserve their characteristics. Additional studies on cross-sections showed that bubbles were formed much deeper (70–100 nm) than the heavily damaged surface layer (10–20 nm), raising concern about the impact on the material mechanical properties conservation and potential additional trapping of hydrogen isotopes.

Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.

We have investigated the synergistic effect of a combination of various impurity gases and hydrogen gas on plasma detachment of high temperature plasma, equivalent to scrape-off layer plasma of tokamaks in the GAMMA 10/PDX end region, utilizing an open magnetic field configuration. A small puff of an impurity gas (N2, Ne, Ar, Kr, Xe) in combination with a puff of H2 gas is examined to evaluate their synergistic effect on the formation of detached plasma; the following results are obtained. (i) A combination of N2 and H2 puffs showed a clear decrease of electron density and ion flux; (ii) N2 and H2 puffs form a strong density gradient along the axial direction; and (iii) other noble impurity gases showed an insufficient synergistic effect. The new results indicate the possibility of achieving a reliable divertor operation scheme and the importance of a deeper understanding of the H2 and N2 assisted recombination process.

International audience; The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment,which went into operation in 2015, has been operating since 2017 with anun-cooled modular graphite divertor. This allowed first divertor physicsstudies to be performed at pulse energies up to 80 MJ, as opposed to 4MJ in the first operation phase, where five inboard limiters wereinstalled instead of a divertor. This, and a number of other upgrades tothe device capabilities, allowed extension into regimes of higher plasmadensity, heating power, and performance overall, e.g. setting a newstellarator world record triple product. The paper focuses on the firstphysics studies of how the island divertor works. The plasma heat loadsarrive to a very high degree on the divertor plates, with only minorheat loads seen on other components, in particular baffle structuresbuilt in to aid neutral compression. The strike line shapes andlocations change significantly from one magnetic configuration toanother, in very much the same way that codes had predicted they would.Strike-line widths are as large as 10 cm, and the wetted areas alsolarge, up to about 1.5 m2, which bodes well for futureoperation phases. Peak local heat loads onto the divertor were ingeneral benign and project below the 10 MW m?2 limit of thefuture water-cooled divertor when operated with 10 MW of heating power,with the exception of low-density attached operation in the high-iotaconfiguration. The most notable result was the complete (in all 10divertor units) heat-flux detachment obtained at high-density operationin hydrogen.

The three-dimensional Monte-Carlo impurity transport and plasma surface interaction code ERO2.0 is applied to a full-torus model for the Large Helical Device (LHD). In order to find an optimum experimental condition for effective real-time wall conditioning (boronization) using an Impurity Powder Dropper (IPD), the toroidal and poloidal distribution of the boron flux density on the divertor components and the vacuum vessel are surveyed in various experimental conditions. The source profile of the neutral boron atoms originated from boron powders supplied from the IPD is calculated using the DUSTT code in background plasmas provided by the EMC3-EIRENE code. The simulations using ERO2.0 predict that higher plasma density operation is inappropriate for the effective wall conditioning because of the toroidally localized boron flux density in a closed helical divertor region. The ERO2.0 simulations have successfully revealed an optimum experimental condition for the wall conditioning with the toroidally uniform boron flux density in the closed helical divertor region.

Injection of impurities in the form of sub-millimeter powder grains is performed for the first time in the Large Helical Device (LHD) plasma, employing the Impurity Powder Dropper (IPD) (Nagy et al., 2018), developed and built by PPPL. Controlled amounts of boron (B) and boron nitride (BN) powder are injected into the helical plasma. Visible camera imaging, UV and charge exchange spectroscopy measurements show that the injected impurities effectively penetrate into the plasma in two different magnetic configurations. The prompt effects of the impurities on the plasma are characterized as the injection rate is scanned. The injected impurities provide a supplemental electron source, causing the plasma density to increase, together with the radiated power. Beneficial effects on the confined plasma temperature are observed at low plasma densities, due to an increased efficiency in NBI power absorption. For n e , a v 1 0 19 m − 3 the powder grains penetrate deeper into the plasma, as they can be less effectively deflected by the plasma flow in the divertor leg, which they have to cross first as they are injected from the top of the machine. In this case, the created B ions are observed to move outwards from UV spectroscopy and charge exchange measurements, due to the outwards direction of the radial electric field. This makes low density plasmas a better candidate for powder boronization techniques.

Characteristics of plasma parameters and turbulence in the isotope-mixing and the non-mixing states in hydrogen-deuterium mixture plasmas in the large helical device are discussed. The isotope mixing state is characterized by the uniform isotope ratio profile regardless of the location of the particle source of each species in the isotope mixture plasma. The isotope non-mixing state is identified by the non-uniform isotope ratio profile measured with bulk charge exchange spectroscopy when the beam fueling isotope species differs from the recycling isotope species. The effect of collisionality, T e / T i ratio, sign of density gradient on transition between isotope mixing and non-mixing is discussed. The plasma parameters preferable for the non-mixing state are found to be lower collisionality, higher T e / T i , and negative or zero density gradient (peaked or flat density profile). The time scale of transition from non-mixing to mixing is evaluated by the hydrogen and deuterium pellet injection near the plasma edge and is found to be less than 5 ms, which is much shorter than the particle confinement time. The strong correlation between isotope mixing and turbulence characteristics is observed. This strong correlation suggests the change in turbulence is a strong candidate for the mechanism causing the transition between uniform and non-uniform isotope density ratio profiles.

During the 19th experimental campaign in the Large Helical Device (LHD) in 2017, deuterium (D) plasma experiments were performed for the first time in the LHD. Tritons were generated by the D-D reaction in the core plasma during the plasma discharges. To remove the tritium in plasma facing surfaces for safety in maintenance, hydrogen (H) plasma experiments were performed for one month after the four months of D plasma experiments. In this study, the removal effect of the tritium by H plasma discharges was investigated. Six stainless steel (SUS316L) samples were introduced into a position of the first-wall of the LHD in the last month of the D experiments, which were exposed to the D plasma for one month and then, three samples were continued to exposure to the following H plasma for another one month. The experimental results showed that one-half of the tritium in the surface region of the SUS sample was removed, which was measured by the TIPT (Tritium imaging plate technique). Further, hydrothermal treatments were performed and the results showed that the tritium retention in the bulk of the SUS materials was also reduced. Meanwhile, the carbon content in the near-surface region was also reduced after H plasma experiments, which reduces the possibility of the formation of the C H chemical bond.

We developed a new Mie-scattering ellipsometry system for the analysis of the generation and transport of dusts in large plasma devices, such as nuclear fusion experiment devices. The system is a rotating-compensator type and consists of polarizer and analyzer modules and a 2-D image sensor. A preliminary experiment was carried out using a small dust plasma cubic chamber. Spherical divinylbenzene polymer particles of $2.25~\pm ~0.05~\mu \text{m}$ in diameter with a dispersion degree of 0.1 $\mu \text{m}$ were injected and suspended in an argon plasma. Light intensity data for 1600 pixels on each image taken 30 times per second were summed as a function of compensator azimuth angle, and then, ellipsometric parameters $\Psi $ and $\Delta $ were determined through Fourier analysis and relational equations. By the comparison of the determined $\Psi $ and $\Delta $ values with the calculated ones based on the Mie theory, it can be stated that they agree fairly well with each other and the measurement method is reliable.

A novel fabrication method for a divertor heat removal component with tungsten (W) armour and copper alloy heat sink was newly developed and named as the “Advanced Multi-Step Brazing (AMSB). The basic principle of multi-step brazing is to apply the “advanced brazing technique” repeatedly during the manufacturing process of the single divertor heat removal component. The advanced brazing technique was originally developed by our previous work for joining between oxide dispersion strengthened copper alloy (ODS-Cu) and W with BNi-6 (Ni–11%P) filler material. The applied ODS-Cu was GlidCop® (Cu-0.3wt%Al2O3). One of the possible examples of the AMSB fabrication process can be considered as follows. First, an appropriate cooling flow path channel is processed into a GlidCop® heat sink. Then, the flow path channel is sealed in a leak tightness condition with a lid made by GlidCop®. The leak tightness joint between GlidCop® (GlidCop®/GlidCop®) can be realized by application of the “advanced brazing technique.” Second, in order to facilitate the weldability to connect the other cooling pipe system, the sleeves made by stainless steel (SUS) are jointed on the interface edge of the flow path channel of the GlidCop® heat sink with a leak tightness condition by the “advanced brazing technique.” Finally, W armour is jointed on the GlidCop® heat sink also by the “advanced brazing technique.” In this study, the mechanical strength of the SUS/GlidCop® joint was confirmed by the three-point bending test. Then, AMSB divertor mock-up with leak tightness condition was successfully produced.

Hydrogen isotope exchange behavior on carbon (C)- tungsten (W) mixed material layer on W was studied as a function of gas exposure time and temperature. It was found that hydrogen isotope exchange enhanced the reduction of residual hydrogen effectively compared by only heating. The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K and above 573 K, thermal annealing reduced the total hydrogen isotope retention. It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer.

The joint of the oxide dispersion strengthened copper alloy, ODS-Cu (ODS-Cu/ODS-Cu), was prepared. The applied ODS-Cu was GlidCop® (Cu-0.3 wt%Al2O3). The joint procedures were based on the brazing technique between GlidCop® and tungsten (W) with BNi-6 (Ni-11%P) filler material. This method was performed without compressive load and without any intermediate layer, and was named the advanced brazing technique. In this study, first, the joint GlidCop®/GlidCop® was prepared without any compressive load. The joint was subjected to the three-point bending test, the Vickers hardness test, microstructure observation and the leak tightness test. The yield strength of the joint was approximately one-half of that of the GlidCop® bulk. In addition, the joint was not leak tightness against fluids. Next, the joint was prepared with compressive load of ˜0.54 MPa which was applied in the direction perpendicular to the joint interface. The yield strength of the joint was almost as high as that of the GlidCop® bulk. Furthermore, the joint achieved leak tightness against fluids. The special feature of this joint method is that a limited volume of the GlidCop® surfaces near the bonding interface are melted during the bonding heat treatment phase at 960 °C due to the eutectic reaction of the Cu-P system. Then, after finalizing the solidification, a tight bonding layer are created without any cavities and cracks. This phenomenon seems to resemble a micro scale welding. The present brazing technique can promise fabrication of a high heat flux component with a complex shaped fluid flow path system with leak tightness.

Processes occurring on the plasma-facing walls in the Large Helical Device (LHD) and in-vessel material migration were investigated using a technique of material deposition probes. Ten Si plates were located on the outer side of the first-wall surface in each 36° toroidal section (Nos. 1–10). Mixed-material layers deposited on the probes during the fiscal year 2012 plasma campaign contained carbon, boron, hydrogen, and metals from which the plasma-facing elements were made. Metallic impurities and hydrogen content in deposited layers were examined by Rutherford backscattering spectrometry (RBS) and elastic recoil detection analysis (ERDA), respectively. The cross-sectional observations of the deposited mixed-material layers were performed with the help of a scanning electron microscope equipped with an energy dispersive X-ray (EDX) spectrometer. The results of the combined RBS/ERDA analysis and TEM-EDX examination allow tracing the processes occurring on plasma-facing components during the plasma campaign.

Ultra-long-pulse helium discharge with ion and electron cyclotron heating (ICH + ECH) in the Large Helical Device (LHD) was achieved in a 48 min plasma (ne ∼ 1.2 × 1019 m−3, Ti,e ∼ 2 keV) with an average heating power of 1.2 MW. The temperature of the first-wall surface during discharges remained at nearly room temperature. However, even in ultra-long-pulse helium discharge, the discharge conditions cannot be said to be in a steady-state, because of two major issues interrupting the steady-state condition. One is the “dynamic change of the wall pumping rate” and the other is the “termination of the discharge with the exfoliation of the mixed-material deposition layers.” Microscopic modifications, such as helium radiation damage and the formation of the mixed-material deposition layers composed of C (∼98%) and Fe (∼2%), on the plasma facing components (PFMs) were clarified to possibly influence the major issues.

This paper describes the results of the experiments performed on Tandem Mirror device GAMMA 10/PDX mainly using a new “divertor simulation experimental module (D-module)” installed on one of the end mirror exits which is specially designed to investigate the physics of plasma detachment. The additional ICRF heating in the anchor-cells, connected to both ends of the central-cell, significantly increases the density in the both cells, which attained the generation of the highest particle flux up to 1023 particles/s m2 at the end-mirror exit. H2 and noble gas injection to enhance the radiation cooling in D-module was performed and a remarkable reduction of the electron temperature (from few tens eV to

Two-dimensional (2-D) distribution of impurity line emissions has been measured with 2-D extreme ultraviolet (EUV) spectroscopy in Large Helical Device (LHD) for studying the edge impurity transport in stochastic magnetic field layer with three-dimensional (3-D) structure. The impurity behavior in the vicinity of two X -points at inboard and outboard sides of the toroidal plasma can be separately examined with the 2-D measurement. As a result, it is found that the carbon location changes from inboard to outboard X -points when the plasma axis is shifted from R ax = 3.6 m to 3.75 m. A 3-D simulation with EMC3-EIRENE code agrees with the result at R ax = 3.75 m but disagreed with the result at R ax = 3.60 m. The discrepancy between the measurement and simulation at R ax = 3.60 m is considerably reduced when an effect of neutral hydrogen localized in the inboard side is taken into account, which can modify the density gradient and friction force along the magnetic field.

In LHD, impurity seeding has been performed to enhance the radiative cooling in the edge region. Neon, nitrogen and argon were seeded by gas puffing, and the behaviour of those impurities in the plasma was investigated with the innovative diagnostic method. Two bolometer arrays were used to measure the two-dimensional radiated power distribution. Using the tomographic technique, radiated power distributions on a poloidal plane can be obtained with the high time resolution. During the discharge with neon puff, considerable radiation from the core region was observed, in addition to the strong edge radiation. In spite of the highly radiated power, plasma did not result in the radiation collapse. On the other hand, in the nitrogen-seeded discharge, the strong radiation only from the peripheral region was observed. Different time evolutions of the total radiated power between neon and nitrogen seeded discharges were observed after stopping each impurity puff.

The wall conditioning using radio frequency (RF) plasma was carried out using Ion Cyclotron Range of Frequency (ICRF) heating and/or Electron Cyclotron Resonance Heating (ECRH) with the working gas of helium under the established confinement magnetic field. After sufficient numbers of repetitive wall discharge conditioning (DC, ICDC for ICRF and ECDC for ECRH), the formation of the parabolic electron density profile and the increase of the central ion temperature Ti were observed. There was no difference in the attained central Ti of the NBI discharge just after the conditioning using a similar value of input energy regardless of the applied RF source. Thus, it is concluded that both ICDC and the ECDC are effective for the higher Ti plasma production under the established magnetic field in the Large Helical Device. The effective RF wall conditioning scenarios are also investigated for high Ti plasma production.