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The toroidal distribution of boron deposition on plasma-facing components (PFCs) in boron powder injection using an impurity power dropper (IPD) was investigated by full-torus simulation and observations in a systematic plasma density-scan experiment. The images of the ablation of dropped boron powders observed with a visible CCD camera were consistently explained by the simulations of the ablation positions of the boron powders considering the size distribution. Simulations assuming full-torus boron deposition on the PFCs did not reproduce the observed intensity profile of boron emission lines for higher plasma densities. It indicated that the density of boron deposited on PFCs installed toroidally far from the IPD was low for higher plasma densities due to the change in the ablation positions of the boron powders toward the outboard side. The experimental results verified the previous full-torus simulation of the toroidal distribution of the boron deposition in both lower and higher plasma densities.

For the first time, we have revealed the spatial structure of a detached plasma with electron–ion recombination (EIR) and molecular-activated recombination (MAR) in response to changes in ion temperature (Ti) and gas pressure. This was achieved using high-density sheet plasma generated by a linear divertor simulator (TPDsheet-ICR). The parallel-plate electrodes were positioned above and below the sheet plasma (∼1019 m−3). We varied the Ti from 3 to 7 eV using ion cyclotron resonance heating and investigated the parameters of the detached plasma near the target in the divergent field region. Plasma emission intensities at Balmer series (Hα and Hγ) and Fulcher band wavelengths (610 ± 10 nm) were measured with a high-speed camera equipped with an Arbaa prism. Additionally, a Langmuir probe measured the electron density and temperature of the plasma. Our results show that as the Ti increases, EIR generated near the plasma periphery disappears, while MAR gradually forms near the plasma center during the transition from detached to attached plasma. This transition occurs because the increase in Ti in the detached plasma transfers energy from ions to electrons, raising the high-energy component of the electrons, resulting in EIR disappearance and the onset of MAR.

Three-dimensionally localization of radiative cooling due to nitrogen (N2) seeding for divertor detachment was detected experimentally. Since the localization along some magnetic field lines induces toroidal asymmetry of heat load reduction on divertor plates, it should be avoided for fusion reactors. The three-dimensionally localized structure was extracted using Principal Component Analysis (PCA) from two-dimensional radiation images measured with an InfraRed imaging Video Bolometer (IRVB). By applying PCA to 34 images each in N2 seeded plasmas with toroidally-asymmetric heat load reduction and in neon (Ne) seeded plasmas with toroidally-symmetric heat load reduction, a radiation feature in N2 seeded plasmas was found as one of the principal components (PC). The three-dimensional transport code EMC3-EIRENE indicated that the ionization in one of the divertor legs is enhanced in nitrogen seeding compared with Ne seeding due to the difference in the first ionization energy. The magnetic field lines from the divertor leg were along the extracted radiation structure and were terminated by the divertor where the heat load decreased due to the N2 seeding. These results indicate that three-dimensionally localized structure of radiative cooling was detected experimentally.

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

A reliable technique for creating joints between tungsten (W) and Reduced Activated Ferritic/Martensitic (RAFM) steel with a pure-copper (Cu) interlayer (W/Cu/RAFM steel) has been explored for developing the heat removal component in the baffle and dome of a divertor. The first joint sample of W/Cu/RAFM steel was fabricated by using a brazing technique described in our previous work. In this work, to evaluate the heat removal property, a joint sample was subjected to a steady-state and a repetitive heat loading test with an electron beam facility (ACT2). In the steady-state heat loading test, a step-wised static heat load with ∼0.7, ∼1.0, ∼1.9, ∼2.7, ∼3.6, ∼4.4 and ∼6.0 MW/m2 was applied to the joint sample. The temperatures on the W and RAFM steel showed a linear increasing tendency, as a function of the heat loading value. Also, in the repetitive heat loading test, a 30 s pulse width of 100 cycles was applied with a maximum heat loading value of ∼2.8 and ∼6.0 MW/m2. The obvious temperature excursion, while increasing the heat cycles, was confirmed at ∼6.0 MW/m2. After the heat loading, a major crack in the vicinity of the interface between the W and Cu was confirmed. This crack led to the degradation of the heat removal performance over ∼6.0 MW/m2.

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

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

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

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

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

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

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

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015–2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and β-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013–2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign. Keywords: Divertor, Tritium, Retention, Deposition, Radiation

A tritium imaging plate technique (TIPT) in combination with an electron-probe microscopic analysis (EPMA) were applied to examine tritium (T) retention characteristics in individual dust particles collected in the Joint European Torus with the ITER-like Wall (JET-ILW) after the first campaign in 2011–2012.A lot of carbon (C)-dominated dust particles were found, which would be pre-existing carbon deposits in the JET-C or released carbon particles from the remaining carbon-fiber components in the JET-ILW. Most of T was retained at the surface of and/or in the C-dominated dust particles. The retention in tungsten, beryllium and other metal-dominated dust particles is relatively lower by a factor of 10–100 in comparison with that in the C-dominated particles. Keywords: JET-ILW, Dust, Tritium, Imaging plate, Electron prove microscopic analysis

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

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.

To reveal the triton transport and the tritium migration in a deuterium plasma experiment in the Large Helical Device (LHD), the distribution of the remaining tritium in divertor tiles made of graphite after the first deuterium plasma experimental campaign in 2017 was investigated. In this study, tritium contents in divertor tiles have been measured by using a full-combustion method. The asymmetric tritium retention in divertor tiles located at symmetric positions, which was found in the previous study by the surface tritium measurement using an imaging plate technique, has been confirmed by the results of the full-combustion method. The asymmetry is considered to be attributed to the asymmetric distribution of lost-points of energetic tritons in divertor. A depth profile of remaining tritium in a divertor tile estimated by using a combination of the imaging plate technique and a sputtering treatment shows that the peak of the profile locates at several micro-meters from the surface. This result suggests that the majority of the remaining tritium impinged upon the divertor tile as energetic tritons. In this study, a distribution of lost-points of energetic tritons has been calculated by using a Lorentz orbit following code (LORBIT) with taking into account divertor components. The obtained distribution has been compared with measured tritium distributions on divertor tiles. The measured and calculated distributions are similar to each other, but they are not the same. The difference between them can be attributed to plasma exposures during and after the deuterium plasma campaign.

N2 gas has been seeded in the Large Helical Device (LHD) to reduce the divertor heat load through enhanced radiation. Radiation is observed by two imaging bolometers, viewing the same poloidal cross-section from top and bottom ports, at a location which is 36° toroidally removed from the N2 gas puff nozzle located at the bottom of the machine. During N2 seeding, these measurements both confirm that additional radiation from the outboard side is coming exclusively from the top of the cross-section, indicating up/down asymmetry, which is also reproduced by modelling with EMC3-EIRENE using a half torus model. In addition, a toroidally localized, magnetic field direction-dependent radiation enhancement is observed with N2 seeding, but is not reproducible by the model.

Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.

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 ne,av

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.

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

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

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

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 ∼23eV to ∼2eV as the neutral pressure in the D-module increased. A clear density rollover was observed at ∼2Pa. 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.

A simple method to evaluate the thickness distribution of the deposition layer formed on the first wall is proposed using an innovative measurement concept of the optical reflection coefficient, which is measured as the RGB (red, green, blue) value using a compact color analyzer. Analysis of the samples exposed to plasmas during an experimental campaign shows the relationship between the thickness of the deposition layer and the reflection coefficient, which is followed by the single layer model. The reflection coefficient clearly indicates the thickness of the deposition layer between 10 and 100nm. The reflection coefficients of stainless steel plates on the helically twisted coil in one of the 10 toroidal sections of the vacuum vessel in the Large Helical Device (LHD) are measured. There is almost no deposition layer on the inner side of the torus, however, the deposition layer reaches a thickness of over 100nm on the first wall near the divertor region. On the outer side of the torus, almost the entire area is covered by the deposition layer. Reflection coefficient measurements indicate that approximately 60% of the area on the measured coil can is coated with a deposition layer over 10nm thick, which suggests that this area plays a role in the wall retention. Keywords: Reflection coefficient, Color analyzer, Deposition layer, Wall retention, Helium

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

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

Effect of carbon based mixed-material deposition layer formation on hydrogen isotope retention was studied. The tungsten (W) samples were placed at four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (higher heat load area), and ER (erosion dominated area) during 2013 plasma experimental campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 4000 shots of hydrogen plasma in a 2013 plasma experimental campaign. Most of the sample surface except for ER was covered by a mixed-material deposition layer formed by plasma experimental campaign, which consisted of carbon, but some metal impurities were contained. For ER sample, He bubbles were formed due to long term He discharge cleaning and He plasma experiments during the plasma experimental campaign. The additional 1keV D2+ implantation was performed to evaluated the D retention enhancement by plasma exposure. It was found that both of H and D retentions were clearly increased. In particular, the H retention was controlled by the thickness of the carbon-dominated mixed-material deposition layer, indicating most of the H was trapped by this mixed-material deposition layer. It is concluded that the accumulation of low-Z mixed-material layer on the surface of the first wall is one of key issues for the determination of hydrogen isotope retention in first wall.