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Intersection points of 1 MeV tritons and the plasma facing wall are visualized in the vacuum vessel of Large Helical Device (LHD) with the divertor plates by the virtual-reality (VR) system. The trajectory of the energetic triton is evaluated by means of the collisionless Lorentz orbit code, and the strike point is calculated by the winding number algorithm. The intersection points of the tritons and the plasma facing wall are visualized as spheres with the rendered internal vacuum wall and divertor plates in the VR space, and it is possible to directly observe the strike points distributed on the wall and on the plates. It is found that many tritons intersect with the divertor plates and that some strike on the vacuum vessel. To evaluate the effectiveness of this VR visualization, we ask people to score the effect of experience and the level of understanding after finding the place where many particles attack on the plasma facing wall on the two-dimensional plane or in the VR space. This visualization helps us to determine where the material probes should be placed on the plasma facing wall in the real LHD for experimental analysis of the accumulated tritium on the plasma facing wall. This development of VR visualization will make a significant contribution to the fusion plasma research.

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.

Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.

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.

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 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.

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

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

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

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

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

In the Large Helical Device, the divertor detachment has been attempted by application of resonant magnetic perturbation (RMP) field and Ne gas puffing in electron cyclotron resonance- heated discharges for compatibility of high central electron temperature and low divertor heat load. Two kinds of divertor detachment phases were observed. The first one appeared transiently just after the Ne gas puffing (1st detachment), and the second one appeared steadily in the latter half of the discharge (2nd detachment). Space-resolved extreme ultraviolet spectroscopy revealed that NeVI-NeVIII emissions increased slightly outside the last closed flux surface (LCFS), while NeIX and NeX emissions increased inside the LCFS in the 1st detachment phase. Although in the 1st detachment the divertor heat load was significantly reduced, the central electron temperature also decreased because the Ne ions were penetrated inside the LCFS as a radiation source. In the 2nd detachment phase, NeVI-NeVIII emissions increased outside the LCFS while NeIX and NeX emissions kept low intensity inside the LCFS. In this phase, low divertor heat load and high central electron temperature were obtained simultaneously because the Ne ions were localized outside the LCFS as a radiation source. The profile measurements of Ne emission show that the edge island structure created by the RMP application impacts on the impurity emission distribution, where the peak of the emission shifts radially stepwise as the detachment proceeds.

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.

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

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.

Isotope effects are one of the most important issues for predicting future reactor operations. Large helical device (LHD) is the presently working largest stellarator/helical device using super conducting helical coils. In LHD, deuterium experiments started in 2017. Extensive studies regarding isotope effects on transport have been carried out. In this paper, the results of isotope effect studies in LHD are reported. The systematic studies were performed adjusting operational parameters and nondimensional parameters. In L mode like normal confinement plasma, where internal and edge transport barriers are not formed, the scaling of global energy confinement time (τ E) with operational parameters shows positive mass dependence (M 0.27; where M is effective ion mass) in electron cyclotron heating plasma and no mass dependence (M 0.0) in neutral beam injection heating plasma. The non-negative ion mass dependence is anti-gyro-Bohm scaling. The role of the turbulence in isotope effects was also found by turbulence measurements and gyrokinetic simulation. Better accessibility to electron and ion internal transport barrier (ITB) plasma is found in deuterium (D) plasma than in hydrogen (H). Gyro kinetic non-linear simulation shows reduced ion heat flux due to the larger generation of zonal flow in deuterium plasma. Peaked carbon density profile plays a prominent role in reducing ion energy transport in ITB plasma. This is evident only in plasma with deuterium ions. New findings on the mixing and non-mixing states of D and H particle transports are reported. In the mixing state, ion particle diffusivities are higher than electron particle diffusivities and D and H ion density profiles are almost identical. In the non-mixing state, ion particle diffusivity is much lower than electron diffusivity. Deuterium and hydrogen ion profiles are clearly different. Different turbulence structures were found in the mixing and non-mixing states suggesting different turbulence modes play a role.

Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

The carbon deposition profile on a Directional Material Probe (DMP) installed in the inboard side of the torus in the Large Helical Device (LHD) is investigated using the ERO2.0 code. The experimental result of the carbon deposition profile with short and wide shadows (lower deposition density areas) on the DMP is reasonably explained by the carbons sputtered from the carbon divertor plates installed in the inboard side. The simulation proves that the short and wide shadows are produced by carbons sputtered from the right and left divertor plate arrays, respectively. The experimental carbon deposition profile accumulated in the previous experimental campaign (FY2010) was successfully reproduced by the simulation, which provides detailed understanding of material (carbon) migration in the divertor region in the LHD.

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.

Isotope effects of ECRH plasma in LHD were investigated in detail. A clear difference of transport and turbulence characteristics in H and D plasmas was found in the core region, with normalized radius ρ < 0.8 in high collisionality regime. On the other hand, differences of transport and turbulence were relatively small in low collisionality regime. Power balance analysis and neoclassical calculation showed a reduction of the anomalous contribution to electron and ion transport in D plasma compared with H plasma in the high collisionality regime. In core region, density modulation experiments also showed more reduced particle diffusion in D plasma than in H plasma, in the high collisionality regime. Ion scale turbulence was clearly reduced at ρ < 0.8 in high collisionality regime in D plasma compared with H plasma. The gyrokinetic linear analyses showed that the dominant instability ρ = 0.5 and 0.8 were ion temperature gradient mode (ITG). The linear growth rate of ITG was reduced in D plasma than in H plasma in high collisionality regime. This is due to the lower normalized ITG and density gradient. More hollowed density profile in D plasma is likely to be the key control parameter. Present analyses suggest that anomalous process play a role to make hollower density profiles in D plasma rather than neoclassical process. Electron scale turbulence were also investigated from the measurements and linear gyrokinetic simulations.

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

To understand the fuel retention mechanism, correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011-2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load.

Twisted-tape-induced swirl-flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q = Q 0 exp(t/τ), τ = 6.04 to 23.07 s) and twisted-tape-induced pressure drop was systematically measured for various mass velocities (G = 4115 to 13,656 kg/m2 s), inlet liquid temperatures (T in = 285.88 to 299.09 K), and inlet pressures (P in = 847.45 to 943.29 kPa) using an experimental water loop flow. Measurements were made over a 59.2-mm effective length and three sections (upper, middle, and lower positions), within which four potential taps were spot-welded onto the outer surface of a 6-mm-inner-diameter, 69.6-mm-heated length, 0.4-mm-thickness platinum circular test tube. Type SUS304 twisted tapes with a width w = 5.6 mm, a thickness δ T = 0.6 mm, a total length l = 372 mm, and twist ratios y = 2.39 and 4.45 were employed in this study. The RANS equations (Reynolds Averaged Navier-Stokes Simulation) with a k–e turbulence model for a circular tube 6 mm in diameter and 636 mm in length were numerically solved for heating of water with a heated section 6 mm in diameter and 70 mm in length using the CFD code, under the same conditions as the experimental ones and considering the temperature dependence of the thermo-physical properties concerned. The theoretical values of surface heat flux q on the circular tubes with twisted tapes with twist ratios y of 2.39 and 4.45 were found to be almost in agreement with the corresponding experimental values of heat flux q, with deviations of less than 30% for the range of temperature difference between the average heater inner surface temperature and the liquid bulk mean temperature ΔT L [ = T s,av - T L , T L = (T in + T out )/2] considered in this study. The theoretical values of the local surface temperature T s , local average liquid temperature T f,av , and local liquid pressure drop ΔP x were found to be within almost 15% of the corresponding experimental ones. The thickness of the conductive sub-layer δ CSL and the nondimensional thickness of the conductive sub-layer y + CSL on the circular tubes with various twisted-tape inserts were determined on the basis of numerical solutions for the swirl velocities u sw ranging from 5.23 to 21.18 m/s. Correlations between the conductive sub-layer thickness δ CSL and the nondimensional thickness of the conductive sub-layer y + CSL for twisted-tape-induced swirl-flow heat transfer in a vertical circular tube were derived.

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%.

Multi-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.

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.

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.