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Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011 -2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, "homogeneous " and "directional ", have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011 -2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, "homogeneous " and "directional ", have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011–2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, “homogeneous” and “directional”, have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition., QC 20240531

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011 -2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, "homogeneous " and "directional ", have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011 -2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, "homogeneous " and "directional ", have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

The ergodic layer in the Large Helical Device (LHD) consists of stochastic magnetic fields exhibiting a three-dimensional structure that is intrinsically formed by helical coils. Spectroscopic diagnostics was employed in the extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength ranges to investigate emission lines of carbon impurities in both hydrogen (H) and deuterium (D) plasmas, aiming to elucidate the impact of distinct bulk ions on impurity generation and transport in the edge plasmas of the LHD. The emission intensity of carbon CIII, CIV, CV, and CVI lines is significantly higher in the D plasma compared to the H plasma, indicating a greater sputtering rate of carbon materials in the D plasma, resulting in a higher quantity of carbon impurities originating from the divertor plates. A Doppler profile measurement of the second order of CIV line emission (1548.20 × 2 Å) was attempted using a 3 m normal-incidence VUV spectrometer in the edge plasma at a horizontally elongated plasma position. The flow velocity reaches its maximum value close to the outermost region of the ergodic layer, and the observed flow direction aligns with the friction force in the parallel momentum balance. The flow velocity increases with the electron density in H plasmas, suggesting that the friction force becomes more dominant in the force balance at higher density regimes. This leads to an increase in the impurity flow, which can contribute to the impurity screening. In contrast, the flow velocity in the D plasma is smaller than that in the H plasma. The difference in flow values between D and H plasmas, when the friction force term dominates in the momentum balance, could be attributed to the mass dependence of the thermal velocity of the bulk ions., source:Plasma 2023, 6(2), 308-321; https://doi.org/10.3390/plasma6020021, source:https://doi.org/10.3390/plasma6020021, identifier:0000-0002-1171-8603

Fusion product diagnostics based on four commercially available single-crystal chemical vapor deposition (s-CVD) diamond detectors are installed in the Large Helical Device (LHD) in order to understand energetic ion confinement. Characteristics of s-CVD diamonds were surveyed using alpha and D-T neutron sources. It is found that the energy resolutions of s-CVD diamonds for ∼ 5 MeV alpha particles and 14 MeV neutrons are 1%–3% and ∼ 1.7%, respectively. Moreover, the response of four s-CVD diamond detectors to alpha particles and D-T neutrons is almost identical. The installation positions of the diamond detectors in the vacuum vessel are searched for, based on the loss points of charged fusion products reckoned by Lorentz orbit calculations. Energy- and time-resolved measurement of fusion product flux will progress in further understanding of energetic ion confinement in LHD., source:K. Ogawa et al 2023 JINST 18 P01022, source:https://doi.org/10.1088/1748-0221/18/01/P01022, identifier:0000-0003-4555-1837

A radiative collapse predictor has been developed using a machine-learning model based on high-density plasma experiments in the Large Helical Device (LHD). Concurrently, the physical background of radiative collapse was discussed based on the distinct features extracted by a sparse modeling, which is one of the frameworks of data-driven science. Electron density, CIV and OV line emissions, and electron temperature at the plasma edge have been extracted as the key parameters of radiative collapse. Those parameters are relevant to the physical knowledge that the major cause of radiative collapse is the enhancement of radiative loss by light impurities in the plasma-edge region. Using these four parameters, the likelihood of occurrence of radiative collapse has been estimated. The behavior of plasma at the edge—in particular, the carbon impurities outside the last closed flux surface—has been evaluated using EMC3-EIRENE code for the phase with increasing likelihood, that is, the plasma is getting close to the collapse. It is shown that the radiation caused by the C3+ ion, which corresponds to the CIV emission, is enhanced in the region where electron temperature is around 10 eV., source:https://doi.org/10.1585/pfr.16.2402010

A radiative collapse predictor has been developed using a machine-learning model with high-density plasma experiments in the Large Helical Device (LHD). The model is based on the collapse likelihood, which is quantified by the parameters selected by the sparse modeling, including ne, CIV, OV, and Te,edge. The control system implementing this model has been constructed with a single-board computer to apply this predictor model to the LHD experiment. The controller calculates the collapse likelihood and regulates gas-puff fueling and boosts electron cyclotron resonance heating in real-time. In density ramp-up experiments with hydrogen plasma, high-density plasma has been maintained by the control system while avoiding radiative collapse. This result has shown that the predictor based on the collapse likelihood has the capability to predict a radiative collapse in real-time., source:http://doi.org/10.1585/pfr.17.2402042

Controlled particulate injections from the PPPL impurity powder dropper (IPD) into the Large Helical Device have demonstrated positive effects on the wall conditions on both an intra and inter-shot basis. Injections over a range of densities, input powers, pulse lengths, heating schemes, injection quantities and main ion species show conclusive evidence of improvement to plasma wall conditions. Successful injections are confirmed by both spectroscopic measurements as well as real-time visible camera signals. In 7 s long plasmas the responses include a reduction in wall recycling as well as a reduction in native impurity content as observed over the course of several discharges. For plasmas longer than 40 s, improvements to the recycling rate and increased impurity control are observed in real time as a consequence of the extended particulate injections. These experiments demonstrate the extended applicability of this solid particulate conditioning technique to the control and maintenance of the plasma wall conditions. In addition they are an important initial step in the development of the real-time boronization technique as a supplement to standard conditioning scenarios., source:R. Lunsford et al 2022 Nucl. Fusion 62 086021, source:http://doi.orig/10.1088/1741-4326/ac6ff5, identifier:0000-0003-3588-6801

In state-of-the-art stellarators, turbulence is a major cause of the degradation of plasma confinement. To maximize confinement, which eventually determines the amount of nuclear fusion reactions, turbulent transport needs to be reduced. Here we report the observation of a confinement regime in a stellarator plasma that is characterized by increased confinement and reduced turbulent fluctuations. The transition to this regime is driven by the injection of submillimetric boron powder grains into the plasma. With the line-averaged electron density being kept constant, we observe a substantial increase of stored energy and electron and ion temperatures. At the same time, the amplitude of the plasma turbulent fluctuations is halved. While lower frequency fluctuations are damped, higher frequency modes in the range between 100 and 200 kHz are excited. We have observed this regime for different heating schemes, namely with both electron and ion cyclotron resonant radio frequencies and neutral beams, for both directions of the magnetic field and both hydrogen and deuterium plasmas., source:Nespoli, F., Masuzaki, S., Tanaka, K. et al. Observation of a reduced-turbulence regime with boron powder injection in a stellarator. Nat. Phys. 18, 350–356 (2022). https://doi.org/10.1038/s41567-021-01460-4, source:https://doi.org/10.1038/s41567-021-01460-4, identifier:0000-0001-7644-751X

0000-0003-4555-1837, Fusion product diagnostics based on four commercially available single-crystal chemical vapor deposition (s-CVD) diamond detectors are installed in the Large Helical Device (LHD) in order to understand energetic ion confinement. Characteristics of s-CVD diamonds were surveyed using alpha and D-T neutron sources. It is found that the energy resolutions of s-CVD diamonds for ∼ 5 MeV alpha particles and 14 MeV neutrons are 1%–3% and ∼ 1.7%, respectively. Moreover, the response of four s-CVD diamond detectors to alpha particles and D-T neutrons is almost identical. The installation positions of the diamond detectors in the vacuum vessel are searched for, based on the loss points of charged fusion products reckoned by Lorentz orbit calculations. Energy- and time-resolved measurement of fusion product flux will progress in further understanding of energetic ion confinement in LHD.

0000-0002-1171-8603, The ergodic layer in the Large Helical Device (LHD) consists of stochastic magnetic fields exhibiting a three-dimensional structure that is intrinsically formed by helical coils. Spectroscopic diagnostics was employed in the extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength ranges to investigate emission lines of carbon impurities in both hydrogen (H) and deuterium (D) plasmas, aiming to elucidate the impact of distinct bulk ions on impurity generation and transport in the edge plasmas of the LHD. The emission intensity of carbon CIII, CIV, CV, and CVI lines is significantly higher in the D plasma compared to the H plasma, indicating a greater sputtering rate of carbon materials in the D plasma, resulting in a higher quantity of carbon impurities originating from the divertor plates. A Doppler profile measurement of the second order of CIV line emission (1548.20 × 2 Å) was attempted using a 3 m normal-incidence VUV spectrometer in the edge plasma at a horizontally elongated plasma position. The flow velocity reaches its maximum value close to the outermost region of the ergodic layer, and the observed flow direction aligns with the friction force in the parallel momentum balance. The flow velocity increases with the electron density in H plasmas, suggesting that the friction force becomes more dominant in the force balance at higher density regimes. This leads to an increase in the impurity flow, which can contribute to the impurity screening. In contrast, the flow velocity in the D plasma is smaller than that in the H plasma. The difference in flow values between D and H plasmas, when the friction force term dominates in the momentum balance, could be attributed to the mass dependence of the thermal velocity of the bulk ions.

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., source:https://doi.org/10.1016/j.nme.2022.101294, identifier:0000-0003-1586-1084

A closed helical divertor (CHD) has been designed for efficient particle control in the plasma periphery and for retaining neutral particles and impurity ions in the divertor region. The effect of impurity retention by divertor legs for the CHD configuration is investigated from the viewpoints of neutral impurity transport and force balance of impurity ions along magnetic field lines. A fully three-dimensional neutral particle transport simulation proves that the plasma on the divertor legs is effective for retaining neutral particles/impurities in the CHD region. A one-dimensional impurity ion transport analysis predicts that friction force by plasma flow from the main plasma sweep impurity ions toward the divertor plates even in high neutral density case in which a steep temperature gradient is formed. It shows that the CHD configuration is promising for enhancing LHD plasma performance by effective control of the neutral particles and the impurity ions in the plasma periphery., source:M. Shoji, S. Masuzaki, T. Morisaki, M. Tokitani, R. Sakamoto, H. Yamada, Effect of divertor legs on neutral particle and impurity retention for a closed helical divertor configuration in the Large Helical Device, Journal of Nuclear Materials, Volume 415, Issue 1, Supplement, 2011, Pages S557-S561, ISSN 0022-3115, https://doi.org/10.1016/j.jnucmat.2010.10.082., source:https://doi.org/10.1016/j.jnucmat.2010.10.082

An optimized closed divertor configuration for effective particle control in LHD is proposed from the viewpoints of the distribution of the strike points and neutral particle transport. Calculations of the distribution of the strike points indicate that 50% of the strike points locate in the inboard side of the torus in a standard magnetic configuration (Rax = 3.60 m). The ratio increases to 80% by installing target plates near lower/upper ports. A three-dimensional neutral particle transport simulation shows that installation of closed divertor components with the target plates raise the neutral pressure in the inboard side by more than one order of magnitude compared to that in the present open divertor case. The analysis of the neutral particle transport predicts that enhancement of the neutral pressure becomes moderate in outward shift configurations (Rax > 3.75 m)., source:M. Shoji, M. Kobayashi, S. Masuzaki, T. Watanabe, H. Yamada, A. Komori, Study of the effect of a closed divertor configuration on neutral particle control in the LHD plasma periphery, Journal of Nuclear Materials, Volumes 390–391, 2009, Pages 490-493, ISSN 0022-3115, https://doi.org/10.1016/j.jnucmat.2009.01.054., source:https://doi.org/10.1016/j.jnucmat.2009.01.054

The measured divertor heat flux profiles are compared to the EMC3-EIRENE simulations for two different times of an LHD discharge, corresponding to higher and lower edge temperatures. The relation between the three-dimensional magnetic field structure and the heat flux distributions on the divertor has been analysed. The modelled heat flux for the lower plasma temperature case has a better agreement with the experimental result obtained by the Langmuir probes, which shows a qualitative reproduction of the experimental profile shape. However, the heat flux distribution for the high plasma temperature case shows a different behaviour between the simulation results and the experimental measurements. The detailed analysis of the heat flux distribution for the higher temperature case which has a larger discrepancy has been performed, both quantitatively and qualitatively. The radiation of the eroded impurity from divertor target plates has a minor effect on the heat flux distribution. Non-uniform cross-field transport coefficients are used in the simulations and its impact on the heat flux distributions is discussed for the case of the high plasma temperature., source:Shuyu Dai et al 2017 Plasma Phys. Control. Fusion 59 085013, source:https://doi.org/10.1088/1361-6587/aa7599

FFHR-d1 is a conceptual design of the Large Helical Device-type fusion reactor. Several design optimizations for FFHR-d1 have been conducted under a multipath strategy. The structural design began from a radial build design of the components using a system code analysis and considering a shielding/breeding blanket design. The support structure must be sufficiently rigid to hold a constant geometric position of the radial build components. Additionally, large apertures are required for the maintenance work of in-vessel components. The shape of the support structure was carefully chosen and the analytically determined stress level was within the permissible limit. Thus, the basic design of the structural components of FFHR-d1 was established. To accelerate the design activity and achieve a consistent helical reactor system, the high-temperature superconducting joint-winding, liquid metal divertor, a supplemental helical coil addition, a novel divertor structure, and other challenging options are investigated alongside the basic design. This paper describes the structural design status of FFHR-d1 by focusing on the latest radial build design, construction scheme, and effectiveness of the novel divertor structure., source:https://doi.org/10.1585/pfr.11.2405061, identifier:0000-0003-4210-9859

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (Rax=3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the in board side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors., source:To cite this article: Mamoru Shoji, Masahiro Kobayashi, Suguru Masuzaki, Akio Sagara, Hiroshi Yamada, Akio Komori & LHD Experimental Groups (2009) Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors, Fusion Science and Technology, 56:2, 1001-1008, DOI: 10.13182/FST09-A9041, source:https://doi.org/10.13182/FST09-A9041

An additional deuterium irradiation was performed for long-term samples mounted on the first wall in the large helical device (LHD) to investigate the effect of mild baking (at 368 K) on fuel hydrogen removal from the modified surface. The efficiency of baking strongly depended on the surface conditions. In the erosion dominant area, almost all the retained deuterium was removed by baking, whereas only 17% and 13% of the retained deuterium were removed from the boron and carbon deposition areas, respectively. The results suggest that the erosion dominant area in the first wall can act as a pumping wall during the main discharge, and, in contrast, the fuel hydrogen retention can continuously increase in boron and carbon deposition areas owing to the low removal efficiency of fuel hydrogen given that the deposition continues during the plasma discharge., source:https://doi.org/10.1585/pfr.12.1302048, identifier:0000-0002-7412-3658

The first attempt to characterize the divertor footprint profile in the heliotron device LHD was done, by using a number of Langmuir probes and the multivariable analysis technique. In order to clarify the generation mechanism of the private-side peak on the footprint profile, which has not been reproduced in the modeling study, over 6000 time points were extracted by excluding time points with profile modifications due to already-known reasons. A characterization index ${r}_{2/1}$ was newly defined from the multivariable analysis result, and its dependences on upstream parameters were investigated. As a result, it was found that the footprint profile correlates with the pressure gradient at the edge inside the core region with a fixed beta, suggesting that change of the plasma pressure profile could modify the edge magnetic field structure even if the volume integral of the plasma pressure was constant., source:H Tanaka et al 2018 Plasma Phys. Control. Fusion 60 125001, source:https://doi.org/10.1088/1361-6587/aae2d7, identifier:0000-0002-8025-008X

Recently, a new limiter concept with liquid metal for a helical fusion reactor was proposed. This work present a modeling study to understand the influence of the limiter to the plasma and to find critical parameters for the liquid limiter concept. EMC3-EIRENE code is applied to a limiter configuration of a vertical column formed with plane surfaces at the inboard side of the Large Helical Device (LHD). The calculation results indicate that heat load on the helical divertor plates is significantly reduced when the limiter is inserted into the ergodic layer even without impurities. However, the influence of impurity radiation is too large in some cases. In order to achieve a large reduction of divertor flux and to avoid a large reduction of core electron temperature, parameter scans with limiter position and sputtering yield were performed. The results suggest that the limiter position is a critical parameter for acceptable degradation of the core and sufficient reduction of divertor flux., source:https://doi.org/10.1585/pfr.13.3403034, identifier:0000-0002-9117-5898

Arcing experiments were conducted in the linear plasma device Pilot-PSI, where a pulsed plasma was superimposed to a steady state plasma. The arcing was observed by a fast framing camera, and the sample was analyzed with a transmission electron microscope. Observations of glowing objects released from the sample in response to the arcing and destruction of the fuzzy layer at the edge of the arc trail without significant melting suggested that dust was formed and released from the surface in response to arcing., source:https://doi.org/10.1585/pfr.15.1205061, identifier:0000-0002-2023-9786

The Large Helical Device (LHD) started the deuterium plasma experiment on March 7, 2017. Approximately 6.4 GBq of tritium was produced in the first deuterium plasma experiment campaign and were utilized as tracers for the investigation of release behavior and the balance of tritium in the LHD vacuum vessel. To determine the tritium balance in LHD, the tritium release from the vacuum vessel was continually observed during the plasma experiment period and the vacuum vessel maintenance activities. The tritium exhaust rate was approximately 32.8% at the end of the plasma experiment. After the plasma experiment, the vacuum vessel was ventilated by room air for the maintenance activity and the tritium release from the in-vessel components was observed. The tritium release rate gradually decreased and became constant after four-month in spite of water vapor concentration. It is suggested that the tritium release mechanism from the vacuum vessel is a diffusion-limited process from the bulk. The tritium release amount during the maintenance activity for one year was approximately 5.0%. Considering the decrease of tritium decay for 1.5 years, tritium inventory in LHD was estimated to be approximately 3.66 GBq (57.2%) at the end of maintenance activity., source:https://doi.org/10.1585/pfr.15.1405062, identifier:0000-0001-9941-1958

We report the observation of arcing damage on the diagnostic shutter during the glow discharge wall conditioning in LHD. The diagnostic system has no experience of plasma discharge produced by electron or ion cyclotron resonance heating or neutral beam injection. The arc trails were observed on the aluminum surface but not on the stainless steel although both materials were exposed to the glow discharge with the same duration. The difference in work functions between two materials may be a cause to divide the conditions of arcing ignition., source:https://doi.org/10.1585/pfr.16.1202061, identifier:0000-0001-6090-5010

Divertor pumping was applied to plasma discharges for superior fuel particle control in the Large Helical Device (LHD). The LHD is equipped with two different pumping systems. One is the main pumping system, in which the pumping speed is 260 m3/s in hydrogen. The other pumping system is the divertor pumping system in which the pumping speed is 70 m3/s in hydrogen. Divertor pumping was applied to 40-second long pulse Electron Cyclotron Heating (ECH) discharges to assess the improvement in particle control provided by divertor pumping. The results show that without divertor pumping, the electron density was not controlled by gas puffing using the feedback signal of line-averaged electron density. Then, the plasma confinement deteriorated, finally leading to radiation collapse. On the other hand, with divertor pumping, the density was well-controlled by gas puffing using the feedback signal. The results indicate that divertor pumping is one of the key tools for controlling the particles in fusion plasmas., source:https://doi.org/10.1585/pfr.16.1202014, identifier:0000-0001-5522-3082

Extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength spectra including line emissions released from neon (Ne) ions ranging from low to high charge states observed simultaneously in a single discharge are summarized for contribution to compile a fundamental spectral dataset for the Ne-seeded divertor heat load reduction experiments in Large Helical Device (LHD). NeIX and NeX lines were observed in the EUV wavelength range of 10∼50 Å and NeIII-NeVIII lines were observed in the VUV wavelength range of 400∼1000 Å. The temporal evolutions of the line intensities exhibited different behaviors between the edge emissions of NeIII-NeVIII with the ionization potential, Ei, of 63∼239 eV and the core emission of NeX with Ei of 1362 eV. NeIX with Ei of 1196 eV exhibited a marginal behavior between the edge emission and the core emission., source:https://doi.org/10.1585/pfr.16.2402006, identifier:0000-0002-1171-8603

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., source:https://doi.org/10.1088/1741-4326/ac3cda, identifier:0000-0001-5220-947X

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., source:https://doi.org/10.1007/s12650-021-00805-8

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., source:https://doi.org/10.1585/pfr.17.2403010, identifier:0000-0003-0655-7347

In order to explore the compatibility of good core plasma performance with divertor heat load mitigation, the interaction between cold edge plasma and core plasma transport, including the edge transport barrier (ETB), has been analysed in the divertor detachment discharges of deuterium plasmas in LHD with resonant magnetic perturbation (RMP) field application. The RMP application introduces a widened edge stochastic layer and sharp boundary in the magnetic field structure between the confinement region and the edge stochastic layer. The widened edge stochastic layer enhances impurity radiation and provides stable detachment operation as compared with the case without RMP. It is found that ETB is formed at the confinement boundary at the onset of detachment transition. However, as the detachment deepens, the resistive pressure gradient-driven MHD mode is excited, which degrades the ETB. At the same time, however, the core transport decreases to keep global plasma stored energy (Wp) unchanged, showing clear core-edge coupling. After a gradual increase of density fluctuation during the MHD activity, a spontaneous increase of Wp and the recovery of ETB are observed while the detachment is maintained. Then, the coherent MHD mode ceases and ELM-like bursts appear. In the improved mode, impurity decontamination occurs, and the divertor heat load increases slightly. Key controlling physical processes in the interplay between core and cold edge plasma are discussed. A comparison between deuterium and hydrogen plasmas shows that hydrogen plasmas exhibit similar features to the deuterium ones in terms of density and magnetic fluctuations, impurity decontamination towards higher confinement, etc. But most of the features are modest in the hydrogen plasmas and thus no clear confinement mode transition with clear ETB formation is defined. Better global confinement is obtained in the deuterium plasmas than the hydrogen ones at a higher radiation level., source:Citation M. Kobayashi et al 2022 Nucl. Fusion 62 056006, source:https://doi.org/10.1088/1741-4326/ac42f3, identifier:0000-0002-0990-7093

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., source:https://doi.org/10.1585/pfr.17.2402022, identifier:0000-0002-1171-8603

In LHD, several methods of detachment have been attempted, including impurity gaspuffing [1], and the application of an m/n=1/1 magnetic perturbation [2]. LHD is equipped with an imaging bolometer (IRVB) [3] that views the plasma from an upper port. Two scenarios are shown and compared, Ne puffing and N2 puffing. In the case of Ne puffing, radiation becomes more intense near the helical divertor X-point as the radiation increases. In the case of N2 puffing, a double stripe pattern evolves around the upper helical divertor X-point, which appears to be localized near the gas puff inlet. In addition, probe data also indicates that the drop in divertor flux with N2 is localized, while uniform with Ne., source:41st EPS Conference on Plasma Physics O3.116, source:http://ocs.ciemat.es/EPS2014PAP/pdf/O3.116.pdf, identifier:0000-0002-4446-4603

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., source:https://doi.org/10.1016/j.nme.2020.100848, identifier:0000-0002-4446-4603

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., source:https://doi.org/10.1585/pfr.16.2403004, identifier:0000-0003-0655-7347

Divertor detachment using higher-Z (Kr) and/or lower-Z (Ne) is investigated. Using Kr+Ne superimposed seeding, plasma radiation could be enhanced at the upstream region in the edge plasma with suppression of impurity accumulation toward the core plasma. Here, the pre-seeded Kr emission was drastically enhanced after the subsequent Ne seeding. Moreover, the detachment using Kr+Ne seeding could be stably sustained while the detachment using single species seeding is short-lived. Reduction of edge electron temperature due to Ne seeding can promote the Kr emission at the upstream region. It indicates the availability of multi-species impurity seeding with different cooling rate., source:https://doi.org/10.1585/pfr.15.1402051, identifier:0000-0003-1586-1084

The 3D peripheral plasma and neutral transport code, EMC3-EIRENE was applied to the Heliotron J with a wide and flexible controllability of magnetic configuration. This code requires a three-dimensional (3D) grid with high resolution in the peripheral plasma region to reproduce the fine plasma structure. The grid generation tool, FLARE, was utilized to create the grid in conjunction with the code developed to arbitrarily set the outer boundary of the peripheral grid. After setting up the 3D grid, we carried out the EMC3-EIRENE calculation successfully for the first time in the Heliotron J's standard configuration. In addition, the convergence of the iterative calculation and the effects of different grid resolutions upon the calculation were investigated. A good numerical convergence was obtained, and the influence of resolution was observed in the electron density in the divertor region., source:https://doi.org/10.1585/pfr.14.3403127, identifier:0000-0002-5730-4158

Density control is crucial for maintaining stable confined plasma. Divertor pumping, where neutral particles are compressed and exhausted in the divertor region, was developed for this task for the Large Helical Device. In this study, neutral particle pressure, which is related to recycling, was systematically scanned in the magnetic configuration by changing the magnetic axis position. High neutral particle pressure and compression were obtained in the divertor for a high plasma electron density and the inner magnetic axis configuration. Density control using divertor pumping with gas puffing was applied to electron cyclotron heated plasma in the inner magnetic axis configuration, which provides high neutral particle compression and exhaust in the divertor. Stable plasma density and electron temperature were maintained with divertor pumping. A heat analysis shows that divertor pumping did not affect edge electron heat conductivity, but it led to low electron heat conductivity in the core caused by electron-internal-transport-barrier-like formation., source:G Motojima et al 2022 Phys. Scr. 97 035601, source:https://doi.org/10.1088/1402-4896/ac5269, identifier:0000-0001-5522-3082

A tungsten block is supposed to be used as a divertor armor material on the helical reactor FFHR-d1. On the other hand, material selection of the heat sink and bonding technique between armor and heat sink are currently under investigation. On the material selection, copper alloy has a large advantage for the thermal conductivity, but its material properties such as toughness and thermal conductivity, are dramatically decreased due to the neutron irradiation. However, from the assessment of the neutronics environment on the divertor region of the FFHR-d1, copper alloys could be used for a heat sink especially at the outer divertor. In the ITER case, copper alloy (CuCrZr) pipes are joined by a brazing technique with Nicuman37 filler material. This combination has not been optimized for the FFHR-d1, because the toughness of the CuCrZr at high temperature over 450 °C is dramatically decreased with increasing the temperature. As such, another candidate is an oxide dispersion-strengthened copper alloy (ODS-Cu) such as GlidCop®. For the bonding technique, a reliable brazing combination between “two kinds of copper alloys” and “three kinds of filler materials (MBF-20, BNi-6, Nicuman37)” were investigated from a viewpoint of mechanical strength. The most superior fracture strength among the three filler materials was BNi-6 with GlidCop®., source:https://doi.org/10.1585/pfr.10.3405035, identifier:0000-0002-3744-2481

An innovative color measurement technique is employed in the Large Helical Device (LHD). This study provides a method for obtaining in broad spatial extent and in great detail the color information of the first wall relating to the thickness of the deposition layer. The RGB (Red, Green, and Blue) value, mainly of the stainless steel plates on the helically twisted coil, is measured by a color analyzer equipped with an integrated sphere light source. On the outer torus side, the colors of almost all stainless steel plates are close to carbon black, which suggests that deposition is dominant. On the inner torus side, all plates except for those neighboring the carbon divertor plates are almost white, as in the case of the stainless steel substrate of the first wall, which suggests that erosion is dominant. The relationship between the color and the distance from the stainless steel plates to the plasma is investigated., source:https://doi.org/10.1585/pfr.10.1202074, identifier:0000-0001-5522-3082

Detached (from the divertors) plasma regime is a very important operational phase for future fusion device operation and is sustained in the Large Helical Device (LHD) by applying an m/n = 1/1 resonant magnetic perturbation (RMP). The radiation from the plasma is found to be localized near the X-point of this RMP during plasma detachment which is confirmed both by the EMC3-EIRENE edge impurity transport code and experiments. This agreement motivates the establishment of a comparison between the results from the transport model and InfraRed imaging Video Bolometer (IRVB) experiments. A synthetic instrument has been developed which can simulate the radiation by tracing the sight lines through the three dimensional carbon impurity radiation results of the EMC3-EIRENE, giving a synthetic image. These synthetic images can be compared with the IRVB image obtained experimentally. Such a comparison is attempted here for two IRVBs installed on LHD for the data from the 2012 and 2013 experimental campaigns. The comparison reveals a fair amount of qualitative agreement but quantitative disagreement among both the approaches. Probable causes for the disagreement are discussed and corrective measures are suggested., source:https://doi.org/10.1585/pfr.9.3402064

Remaining tritium in the vacuum vessel after the first deuterium plasma experimental campaign conducted over four months was investigated in the large helical device (LHD) for the first time in stellarator/heliotron devices by using the tritium imaging plate technique. In-vessel components such as divertor tiles and first wall panels, and long-term material probes retrieved from the vacuum vessel were analyzed. The in-vessel component in which tritium remained most densely is the baffle part of divertor tiles made of graphite retrieved from the inboard-side divertor. Asymmetric tritium retention is observed on divertor tiles located at magnetically symmetric positions, and can be attributed to the toroidal field direction dependence of the asymmetric loss of energetic tritons generated by deuterium–deuterium nuclear fusion reactions. On the first wall, tritium remained in a deposited layer, which mainly consists of carbon., source:S Masuzaki et al 2020 Phys. Scr. 2020 014068, source:https://doi.org/10.1088/1402-4896/ab6d44, identifier:0000-0003-0161-0938

The divertor strike line pattern on the helical divertor of LHD was observed with an infra red camera. The derived heat flux pattern show multiple distinct strike lines depending on the equilibrium magnetic configuration. Predictions of such divertor heat loads thus require a modeling of the magnetic configuration and the heat transport in the magnetic edge. Equilibrium magnetic topologies were analyzed with HINT2, while the plasma fluid model code EMC3 was used to simulate the energy transport in the edge. The measured multi peak structure of the divertor heat flux is correlated to the intersection points of elongated loop shaped flux tubes of long LC field lines. But the fluid model could not recreate the total energy load and the multiple heat flux peaks on the divertor. A Variation in the plasma density ne as a transport parameter in order to fit the simulated heat flux to the measured one shows a contradicting tendency., source:https://doi.org/10.1585/pfr.8.2402126

The key to improving the heat load of vacuum plasma sprayed tungsten coatings on low activation ferritic steel maintained at low temperatures is elimination of stratified low-density layers with many large pores, in which thermal cracks propagate preferentially. The low-density layers are formed owing to the deposition of large solidified tungsten particles, which remain mainly at the periphery of the spray stream. In this study, by shading this periphery, partially homogeneous tungsten coatings without large pores were successfully obtained. The coatings are expected to show good heat load, which is feasible for nuclear fusion applications., source:https://doi.org/10.1585/pfr.8.1405111

Distributions of deposits and hydrogen (H) on the graphite divertor target elements TM4h4 and TM3v5 in the test divertor units 3 (TDUs3) of Wendelstein 7-X (W7-X) are studied. The TM4h4 and TM3v5 are located at the magnetically symmetric positions in the upper and lower divertor. The microstructure of the deposition layer is characterized by a transmission electron microscope (TEM) combined with a focused ion beam (FIB). Metallic deposits such as iron (Fe), molybdenum (Mo), chromium (Cr) are detected in the deposition layer by energy-dispersive x-ray spectroscopy (EDS). The depth-resolved distribution patterns of boron (B) and metallic deposits on upper and lower horizontal (h) divertor target elements TDUs3-TM4h4 as well as upper and lower vertical (v) divertor target elements TDUs3-TM3v5 are clarified by glow discharge optical emission spectrometry (GDOES). Results for both TDUs3-TM4h4 and TDUs3-TM3v5 show that the B deposition regions exhibit higher H retention due to the co-deposition with deposits. On the other hand, up-down asymmetries in B deposition caused by particle drift exist on both TDUs3-TM4h4 and TDUs3-TM3v5. The B deposition amount on upper TDUs3-TM4h4 is 40% smaller than that on lower TDUs3-TM4h4. While for the vertical target elements, the B deposition amount on upper TDUs3-TM3v5 is 35% larger than that on lower TDUs3-TM3v5. Meanwhile, a shift of around 3 cm in B deposition peaks is observed on upper and lower TDUs3-TM4h4 and TDUs3-TM3v5. Results of numerical simulation of carbon deposition/erosion profiles on the target elements using ERO2.0 code and power flux measured by infrared cameras are shown and compared with the above mentioned B profiles., source:Mingzhong Zhao et al 2022 Nucl. Fusion 62 106023, source:https://doi.org/10.1088/1741-4326/ac8c56, identifier:0000-0001-5089-3642

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., source:https://doi.org/10.1016/j.nme.2022.101257, identifier:0000-0003-0655-7347

In order to study migration of sputtered materials and profiles of their re-deposition in the Large Helical Device (LHD), material probes fixed with Si specimens were installed on the private region of the torus outer first-wall in all toroidal sections (10 sections) in the experimental campaign in 2011. The material probes fixed with the Si specimens were exposed to various wall conditionings such as glow discharges cleanings (GDCs), boronization, Ti sublimation and main plasma discharges. After the campaign, the specimens surfaces were investigated with visual observation, Scanning Electron Microscope (SEM) and Energy Dispersive X-ray Spectroscopy (EDS). A type of visual deposition (band) pattern has been observed on the probes at almost all toroidal sections of LHD. The analysis of these patterns on the material probes informs the source and direction of material migration inside the LHD vacuum vessel. The observation and result of the experiment have been presented in this paper., source:https://doi.org/10.1585/pfr.8.2402074

A joint of oxide dispersion strengthened copper alloy (ODS-Cu; GlidCop®), i.e., GlidCop®/GlidCop® joint and a joint of stainless steel (SUS) and GlidCop®, i.e., SUS/GlidCop® joint were prepared by using the advanced brazing technique (ABT). Deformation and fracture behavior, and thermal stability of these joints were investigated by means of three-point bending test, scanning electron microscope (SEM) and energy dispersive X-ray spectroscopy (EDS). Firstly, in the case of the GlidCop®/GlidCop® joint, plastic deformation occurred predominantly in the diffusion layer with lower hardness. The fracture proceeded approximately at the center of the diffusion layer. In the case of the SUS/GlidCop® joint, on the other hand, plastic deformation occurred widely in the bulks, GlidCop® and SUS near the interfaces between the bulk and the bonding layer. The fracture proceeded mainly along the interface between the bonding layer and the bulk GlidCop®, and substantially along the interface between the bonding layer and the bulk SUS. Secondly, influences of 2nd-step heat on the GlidCop®/GlidCop® and SUS/GlidCop® joints were examined from viewpoints of microstructure and mechanical strength. Almost no obvious changes in the width of the diffusion layer (GlidCop®/GlidCop® joint) or the bonding layer (SUS/GlidCop® joint), and in the element distribution in these layers after the 2nd-step heat were recognized. In addition, any obvious changes in the shape of stress-strain curve and yield strength were not recognized for both joints., source:https://doi.org/10.1016/j.fusengdes.2022.113312, identifier:0000-0002-3744-2481

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., source:https://doi.org/10.1016/j.nme.2022.101294, identifier:0000-0003-1586-1084

0000-0001-5089-3642, Distributions of deposits and hydrogen (H) on the graphite divertor target elements TM4h4 and TM3v5 in the test divertor units 3 (TDUs3) of Wendelstein 7-X (W7-X) are studied. The TM4h4 and TM3v5 are located at the magnetically symmetric positions in the upper and lower divertor. The microstructure of the deposition layer is characterized by a transmission electron microscope (TEM) combined with a focused ion beam (FIB). Metallic deposits such as iron (Fe), molybdenum (Mo), chromium (Cr) are detected in the deposition layer by energy-dispersive x-ray spectroscopy (EDS). The depth-resolved distribution patterns of boron (B) and metallic deposits on upper and lower horizontal (h) divertor target elements TDUs3-TM4h4 as well as upper and lower vertical (v) divertor target elements TDUs3-TM3v5 are clarified by glow discharge optical emission spectrometry (GDOES). Results for both TDUs3-TM4h4 and TDUs3-TM3v5 show that the B deposition regions exhibit higher H retention due to the co-deposition with deposits. On the other hand, up-down asymmetries in B deposition caused by particle drift exist on both TDUs3-TM4h4 and TDUs3-TM3v5. The B deposition amount on upper TDUs3-TM4h4 is 40% smaller than that on lower TDUs3-TM4h4. While for the vertical target elements, the B deposition amount on upper TDUs3-TM3v5 is 35% larger than that on lower TDUs3-TM3v5. Meanwhile, a shift of around 3 cm in B deposition peaks is observed on upper and lower TDUs3-TM4h4 and TDUs3-TM3v5. Results of numerical simulation of carbon deposition/erosion profiles on the target elements using ERO2.0 code and power flux measured by infrared cameras are shown and compared with the above mentioned B profiles.