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Neutron transport calculations with a three-dimensional model of the helical reactor FFHR-d1 have been performed for the accurate analysis of neutronics environment in the divertor areas. Based on the obtained neutron spectra, magnitudes of irradiation damage, contact dose rates and decay heat have been evaluated mainly for Fe, W and Cu. Since divertors can be placed behind radiation shields in helical reactors, magnitudes of damage and radioactivation at the outboard divertors are almost two orders lower than those at blanket first walls. Cu materials could be used as a cooling channel material of the outboard divertors. In contrast, the magnitudes at the inboard divertors are only one order lower compared with those at the first walls due to the limited space at the inboard side. Damage on Cu is evaluated to be ∼10 dpa after 6 years operation. Further efforts in divertor development and reactor design could suppress the magnitude of damage to less than half for adoption of Cu materials for the inboard divertors.

To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tungsten samples were placed at three or four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) during 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ∼5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0 keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DP sample was derived compared to that for pure W. In especially, the carbon-dominated mixed-material layer would control the hydrogen isotope retention for all the area except for erosion-dominated area, indicating that the chemical structure for carbon-dominant mixed-material layer would govern the H and D retention enhancement for most area by long-term plasma exposure. Therefore, the surface area for carbon material would be one of key issues for the determination of hydrogen isotope retention in first wall, even if all tungsten first walls will be used.

FFHR-d1 is a conceptual design of the helical reactor being developed at the National Institute for Fusion Science. The maintenance of in-vessel components is very important for the fusion demo reactor. In addition, sufficient pathways are needed for the divertor exhaust. To implement these, the vacuum vessel, coil support structure, and cryostat require large apertures. However, the coil support structure has to be sufficiently rigid to remain within soundness and deformation limits. A design combining the structural components in the FFHR-d1A was developed from mechanical and thermal viewpoints. Consequently, components having a sufficiently large port area were provided. An investigation of the maintenance and exhaust schemes has been planned on the basis of this fundamental design.

In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m−2 at the test divertor units (TDU), ≤500 kW m−2 at the baffles, heat shields and toroidal closures and ≤100 kW m−2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

An extended mesh system for EMC3-EIRENE has been developed to simulate peripheral plasma including the ergodic and the divertor leg regions of LHD. Both the open and the closed divertor configurations are available. A series of simulations for 8MW input power, five different electron densities at the LCFS (last closed flux surface) and the open/closed configurations were carried out. Approximately 10 times larger neutral pressure was observed under the dome structure compared with the open configuration, which is in good agreement with experimental measurements. In the case of the closed configuration, the leg regions have a large contribution of ionization to hydrogen recycling. In the case of high density discharges, however, electron temperature in the legs becomes low and the major contribution of ionization moves to the ergodic region. Significant influence of configurations is observed in the inboard side of LHD, where closed divertor components are installed but little influence is seen near the LCFS. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Measurement and control of divertor heat flux are a crucial issue for realization of fusion burning plasma. However, time-dependent heat flux is often difficult to be estimated experimentally, because thermal diffusion process in thermal probe tips takes time longer than the discharge duration of present experiments. A new reconstruction method for time dependent heat flux from the measurement of the thermal probe tip temperature is proposed in this paper. The heat flux of the Large Helical Device divertor leg was analyzed by the proposed method as the function of the time for the first time. This indicates that heat flux measurement with thermal probes would become a monitoring tools even for the future fusion devices. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

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.

Ion sensitive probe (ISP) enables simultaneous measurement of ion temperature, electron temperature, and density. So far, we have observed a difference between electron temperature, and density data obtained using ISP and Langmuir probe (LP) in divertor experimental module of GAMMA 10/PDX. To clarify the difference observed, in this study comparisons of two kinds of ISPs are performed. In Type-I ISP, in which guard electrode is aligned with outer dielectric tube, shows an influence of temperature anisotropy on the electron temperature measurement. The electron density detected using Type-I ISP is considerably smaller than the one detected using LP. In the case of Type-II ISP, in which the guard electrode protrudes from the outer dielectric tube, the electron temperature and density detected are similar to the data obtained using LP. Furthermore, the discovered effects of hydrogen (H2) gas pressure change are presented here.

Toroidal profiles of the microscopic damage and erosion/deposition on the Large Helical Device (LHD) first-wall (SUS316L) were simultaneously evaluated. 10 pairs of SUS316L and Si specimens were mounted on 10 sets of special holders, and were located on the outer side of the first-wall surface in each 36° toroidal angle section (Nos. 1–10). For separate determination of the effects of glow discharge cleanings (GDCs) and main plasma discharges, two types of holders —“floating-potential” and “ground-potential”— were used in each toroidal section. The former was electrically insulated from the first wall; therefore, energetic ions could not be injected into the specimens during GDC. Hence, we could analyze two cases: with and without GDCs exposure. Sputtering erosion of the first-wall surfaces was mainly caused by GDCs and not main plasma discharges, and the erosion depths of each toroidal section were varied from 50 nm to 1 μm. Characteristics of the deposition layers and microscopic damages on the SUS316L matrix were different in each toroidal section.

A conceptual design for the helical DEMO reactor FFHR-d1 is under development at the National Institute for Fusion Science. In the large helical device (LHD)-type configuration of FFHR-d1, the space between the plasma boundary and the helical coil is limited. Many components are required to be installed at definite geometric positions within this space. The position of each component changes owing to variations in the surrounding conditions, such as temperature, electromagnetic forces, and support methods. This paper proposes a radial-build design concept; the application of this concept ensures constant geometric positions of the components and the gaps between them during normal operation and the construction/maintenance phase. On the other hand, large apertures are required for the coil support structure for the maintenance of blanket and divertor systems. This paper presents a structural design procedure for the coil support structure with large apertures and an LHD-type gravity support post. The analytic results show that the stress level of the structure is lower than the permissible limit.

The twisted-tape-induced swirl flow heat transfer due to exponentially increasing heat inputs with various exponential periods and twisted-tape-induced pressure drop were systematically measured with mass velocity G = 4120 to 13 570 kg/m2/s, inlet liquid temperature Tin = 300.13 to 305.78 K and inlet pressure Pin = 866.52 to 945.86 kPa by an experimental water loop flow. Measurements were made on a 59.2 mm effective length and its three sections (upper, mid and lower positions), which were spot-welded four potential taps on the outer surface of a 6 mm inner diameter, a 69.6 mm heated length and a 0.4 mm thickness of Platinum circular test tube with the twisted-tape insert. The SUS304 twisted-tape of width w = 5.6 mm, thickness δT = 0.6 mm, total length l = 372 mm, pitch of 180° rotation H = 20.34 mm and twist ratio y = H/d = 3.39 was employed in this work. On the other hand, theoretical equations for k–ɛ turbulence model in a circular tube of a 6 mm in diameter and a 636 mm long with the twisted-tape insert were numerically solved for heating of water with heated section of a 6 mm in diameter and a 70 mm long by using PHOENICS code under the same conditions as the experimental ones considering the temperature dependence of thermo-physical properties concerned. The twisted-tape of w = 5.6 mm, δT = 0.6 mm, l = 370 mm, H = 20 mm and y = 3.33 was installed under the same experimental position. The surface heat flux q and the average surface temperature Ts,av on the circular tube with the twisted-tape of y = 3.33 obtained theoretically were compared with the corresponding experimental values on q versus the temperature difference between average heater inner surface temperature and liquid bulk mean temperature ΔTL [=Ts,av − TL, TL = (Tin + Tout)/2] graph. The numerical solutions of q and ΔTL are almost in good agreement with the corresponding experimental values of q and ΔTL with the deviations less than 0% to +20% for the range of ΔTL tested here. The numerical solutions of the local surface temperature (Ts)z, local average liquid temperature (Tf,av)z and local liquid pressure drop ΔPz were also compared with the corresponding experimental data of (Ts)z, (Tf,av)z and ΔPz versus heated length L or distance from inlet of the test section Z graph, respectively. The numerical solutions of (Ts)z, (Tf,av)z and ΔPz are within ±5% difference of the corresponding experimental data on (Ts)z, (Tf,av)z and ΔPz. The thickness of the conductive sub-layer δCSL [=(Δr)out/2] and the non-dimensional thickness of the conductive sub-layer y+CSL [=(fF/2)0.5ρluδCSL/μl] for the turbulent heat transfer on the circular tube with the twisted-tape insert are clarified based on the numerical solutions at the swirl velocity usw ranging from 5.39 to 18.03 m/s. The correlations of δCSL and y+CSL for twisted-tape-induced swirl flow heat transfer in a vertical circular tube are derived.

The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages.

This paper describes the results of characterization of high heat and particle fluxes produced at the end-cell of the large tandem mirror GAMMA 10 and of the initial plasma-irradiation experiments. In the case of ICRF plasmas, the heat flux of 0.8 MW/m and the particle flux of 4 × 10/m s were achieved at the end-mirror exit. The heat flux increases with the ICRF power and has a linear relationship with the stored energy. Direct ion energy analysis clarified that the parallel ion temperature can be controlled from 100 eV to 400 eV by changing the ICRF power. Additional plasma heating using another ICRF system in the anchor-cell significantly increases the particle flux, which gives a clear prospect of generating the higher particle flux by applying additional ICRF heating in the neighboring cells. The initial results of the plasma-gas-material interactions on a new V-shaped tungsten target were also reported. © 2013 Elsevier B.V. All rights reserved.

Tungsten (W) samples, undamaged and damaged by irradiation with 2.4 MeV Cu2+, were placed flush to the first wall of Large Helical Device (LHD) nearby the divertor target, and were exposed to 134 shots of hydrogen plasma discharges. Thereafter, to evaluate the enhancement of hydrogen isotope retention for W due to LHD hydrogen (H) plasma exposures, the W samples with and without LHD H plasma exposures were implanted with 1.0 keV D 2 + . For W samples without LHD H plasma exposures, the D retention in the damaged W was by a factor of ∼6 higher than that in the undamaged W. After LHD H plasma exposures, the carbon layers ∼4 nm in thickness were deposited on the W surfaces. Due to trapping of D atoms in the carbon layers, the D retention enhancement for undamaged and damaged W samples exposed to LHD H plasmas was significantly higher than that for the W samples without LHD H plasma exposures.

In fusion devices, dust particles are generated due to plasma–wall interactions and may cause safety or operational problems. Therefore it is necessary to clarify the generation and transport mechanisms of dust particles. Here we have measured energy influx from H 2 plasmas toward a graphite target using a calorimetric probe and compared the results with the dust flux toward a dust collecting substrate set on the reactor wall. The dust flux decreases with increasing the energy influx. For the higher discharge power, the more number of dust particles tend to redeposit onto the graphite target due to the higher ion drag force and hence the dust flux toward the reactor wall becomes smaller. The results show that dust inventory depends strongly on energy influx to graphite divertor plates in fusion devices.

Fusion plasma research crucially needs to understand the characteristics of the deposition layer that forms in the vacuum vessel because of the migration of plasma-facing materials. Such a deposition layer can accumulate on the fuel particles on account of a co-deposition process and can be a source of dust in the plasma. In this study, nano-geological diagnosis which uses a focused ion beam (FIB) fabrication technique and transmission electron microscope (TEM) observations, has been applied to the mixed-material deposition layer formed in a large helical device (LHD). A cross-sectional view of the deposition layer with nanometer resolution was successfully observed. The total thickness of the layer was estimated at ∼8 μm, and it has very fine and stratified layer structures in the nanometer level. The characteristics of each layer are indicative of the operational history of the LHD and the mechanism of dust formation by layer flaking.

Spatial profiles of plasma flow and Mach number in the stochastic magnetic boundary layer as well as ion temperature ( T i ) and electron temperature ( T e ) in the divertor region in Large Helical Device (LHD) have been studied by a movable multiple functions probe, which consists of Mach probes and an ion sensitive probe. The results of ion saturation current measurements indicated plasma flow direction is alternated in the stochastic magnetic boundary. Mach number profiles for different plasma densities have been evaluated experimentally which compared with 3-D transport code. T i and T e in the divertor region measured by the ion sensitive probe decreased with increasing line-averaged density. Although T i was higher than T e in the low density plasma, both temperatures became almost the same at higher density.

Deuterium retention in isotropic graphite (ETU-10) was investigated by using nuclear reaction analysis (NRA) under high flux steady state plasma exposure, relevant to divertor plasma conditions. The sample temperature was controlled precisely from 350 K to 550 K with a step of 50 K by using air cooling and electron heating. The sample was irradiated by deuterium plasma at three different fluxes up to 1.0 × 10 18 cm −2 s −1 with the incident ion energy of 33 eV. As the result, dynamic retention was clearly observed. The dynamic retention decreased with the sample temperature and increased with the plasma flux (at the flux of 1.0 × 10 18 cm −2 s −1 , dynamic retention was 8.8 × 10 16 cm −2 at 350 K and 3.5 × 10 16 cm −2 at 550 K). However, a fraction of the dynamic retention normalized by total retention for each sample temperature remained constant. Furthermore, the sample temperature dependence of release time was observed (48–84 ms).

In the Large Helical Device (LHD), studies of divertor heat and particle control have been conducted with a helical divertor magnetic structure which exists naturally in the heliotron-type magnetic configuration to explore operation scenarios in a heliotron-type fusion reactor. Reduction of the divertor heat load has been one of the focused studies in the recent LHD experiment. Impurity seeding and modification of the edge magnetic structure with a resonant magnetic perturbation field have been investigated to enhance radiation power. To control the fueled particles and impurities, the helical diveror will be closed with a baffle structure and in-vessel cryo-sorption pump in 2012. For the examination of the baffle structure, it was partially installed in LHD in 2010. Comparison of the neutral particle pressure in the divertor with the baffle structure and with the existing open divertor shows that the baffle structure efficiently compresses fueled particles in the divertor.

We measured the dependence of the collected flux of dust particles on the bias potential Δ ϕ with respect to ground in the Large Helical Device at the National Institute of Fusion Science. This is a first step toward developing a novel method for removing dust for fusion reactors based on applying a local bias potential. The collected dust particles were classified into two kinds: spherical carbon particles smaller than 300 nm and stainless-steel flakes smaller than 1 μm. For Δ ϕ = −70 to +70 V, the flux of spherical particles increased exponentially with increasing Δ ϕ . The flux of flakes increased when Δ ϕ was increased from −70 V to +30 V and it decreased at higher Δ ϕ . These results suggest that this method is useful for removing carbon and metal dust particles from the shadow area of fusion devices using appropriate control of the local bias potential.

The steady-state turbulent heat transfer (THT) due to exponentially increasing heat inputs with various exponential periods (Q = Q0exp(t/τ), τ = 6.55–21.81 s) were systematically measured with the flow velocities, u, of 4.15, 7.05, 10.07 and 13.50 m/s by an experimental water loop flow. Measurements were made on a 6 mm inner diameter, a 59.2 mm effective length and a 0.4 mm thickness of HORIZONTAL Platinum (Pt) circular test tube. The relation between the steady-state turbulent heat transfer and the flow velocity was clarified. The steady state nucleate boiling heat transfer (NBHT) and the steady state critical heat fluxes (CHFs) of the subcooled water flow boiling for HORIZONTAL SUS304 circular test tube were systematically measured with the flow velocities (u = 3.93–13.86 m/s), the outlet and inlet subcoolings (ΔTsub,out = 60.40–130.30 K and ΔTsub,in = 81.30–154.20 K), the inlet pressures (Pin = 786.29–960.93 kPa) and the increasing heat input (Q0exp(t/τ), τ = 8.36 s). The HORIZONTAL SUS304 test tube of inner diameter (d = 6 mm), heated length (L = 59.4 mm), effective length (Leff = 48.4 mm), L/d (=9.9), Leff/d (=8.06) and wall thickness (δ = 0.5 mm) with surface roughness (Ra = 3.89 μm) was used in this work. The NBHT and the steady state CHFs of the subcooled water flow boiling for the HORIZONTAL SUS304 circular test tube were clarified at the flow velocities, u, ranging from 3.93–13.86 m/s. The steady-state THT data, the NBHT ones and the steady state CHF ones were compared with the values calculated by authors’ THT correlation, their NBHT ones and their transient CHF ones against outlet and inlet subcoolings based on the experimental data for the VERTICAL circular test tubes with the flow velocities, u, ranging from 4.0 to 42.4 m/s. The influences of test tube orientation on the THT, the NBHT and the subcooled flow boiling CHF are investigated into details and the widely and precisely predictable correlations of the THT, the NBHT and the transient CHFs against outlet and inlet subcoolings in a HORIZONTAL circular test tube are derived based on the experimental data. The THT correlation, the NBHT ones and the transient CHF ones for the HORIZONTAL test tube can describe the THT data, the NBHT ones and the subcooled flow boiling CHF ones for the wide ranges of ΔTsub,out, ΔTsub,in and u obtained in this work within ±15% difference.

Control of tritium retention and its removal from the first wall of future fusion devices are the most crucial issues for safety and effective use of the fuel. Nitrogen injection into edge plasmas has been considered and tested as an effective method for suppression of carbon dust growth and reduction of hydrogen isotope inventory. In this paper we have investigated scavenger effects of nitrogen injected into H2/CH4 plasmas using a small helical device where low density (ne ∼ 1016 m-3) and low temperature (Te = 5-10 eV) hydrogen plasmas are generated in steady state condition like remote plasmas in fusion devices. It is shown from the comparison of the carbon film deposition and particle growth between those with and without N2 injection that the chemical erosion effects of nitrogen gas on the carbon film and particle growth strongly depends on the surface temperature. With increasing the surface temperature higher than ∼400 K, the nitrogen chemical erosion significantly works to reduce the hydrocarbon deposition. © 2013 Elsevier B.V. All rights reserved.

Divertor simulation study has been started as a new research plan, by making best use of a large linear plasma device. The experiment of generating the plasma flow with high heat and particle flux was successfully performed at an end-mirror exit of the GAMMA 10 tandem mirror. In typical hot-ion-mode plasmas, the heat-flux density of 0.6 MW/m2 and the particle-flux density of 1022 particles/s m2 were simultaneously achieved in the case of only ICRF heating and superimposing the 300 kW ECH pulse attained the peak value of the net heat-flux up to 8 MW/m2 on axis. The above experimental results and the simulation analysis of ICRF heating using the Fokker-Planck code give a clear prospect of generating the required performance for divertor studies by building up the plasma heating systems to the end-mirror cell. Detailed behavior of the plasma flow and the future research plan are also described. © 2010 Elsevier B.V. All rights reserved.

We have investigated plasma detachment phenomena of high-density helium plasmas in the linear plasma device Pilot-PSI, which can realize a relevant ITER SOL/Divertor plasma condition. The experiment clearly indicated plasma detachment features such as drops in the plasma pressure and particle flux along the magnetic field lines that were observed under the condition of high neutral pressure; a feature of flux drop was parameterized using the degree of detachment (DOD) index. Fundamental plasma parameters such as electron temperature (Te) and electron density in the detached recombining plasmas were measured by different methods: reciprocating electrostatic probes, Thomson scattering (TS), and optical emission spectroscopy (OES). The Te measured using single and double probes corresponded to the TS measurement. No anomalies in the single probe I – Vcharacteristics, observed in other linear plasma devices [ 16, 17,36], appeared under the present condition in the Pilot-PSI device. A possible reason for this difference is discussed by comparing the different linear devices. The OES results are also compared with the simulation results of a collisional radiative (CR) model. Further, we demonstrated more than 90% of parallel particle and heat fluxes were dissipated in a short length of 0.5 m under the high neutral pressure condition in Pilot-PSI.

The twisted-tape-induced swirl flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q = Q0 exp(t/τ), τ = 7, 14 and 23 s) and the twisted-tape-induced pressure drop were systematically measured with mass velocities, G, ranging from 4022 to 15,140 kg/m2 s by an experimental water loop flow. Measurements were made on a 59.2 mm effective length which was spot-welded two potential taps on the outer surface of a 6 mm inner diameter, a 69.6 mm heated length and a 0.4 mm thickness of platinum circular test tube. The twisted tapes with twist ratios, y [=H/d = (pitch of 180° rotation)/d], of 2.39, 3.39 and 4.45 were used in this work. The relation between the swirl velocity and the pump input frequency and that between the fanning friction factor and Reynolds number (Red = 2.04 × 104 to 9.96 × 104) were clarified. The twisted-tape-induced swirl flow heat transfers with y = 2.39, 3.39 and 4.45 were compared with the values calculated by our correlation of the turbulent heat transfer for the empty tube and other worker's one for the circular tube with the twisted-tape insert. The influence of y and Reynolds numbers based on swirl velocity, Resw, on the twisted-tape-induced swirl flow heat transfer was investigated into details and the widely and precisely predictable correlation of the twisted-tape-induced swirl flow heat transfer was derived based on the experimental data. The correlation can describe for the twisted-tape-induced swirl flow heat transfer for the wide ranges of twist ratios (y = 2.39–4.45), mass velocities (G = 4022–15140 kg/m2 s) and Reynolds numbers based on swirl velocity (Resw = 2.88 × 104 to 1.22 × 105) within −10 to +30% difference.

Nano-geological diagnosis has been applied to the deposition layer formed on the Si specimens installed at the first wall surface during 2007FY Large Helical Device (LHD) experiment. Cross-sectional views of the deposition layers with nanometer-resolution were successfully observed by using focused ion beam fabrication technique and transmission electron microscopy observations. On the specimen located close to the divertor plates, deposition layer with ∼800 nm thick was observed, and it has very fine and stratified layer structures in nanometer-level. Thickness of each layer is so various from 5 to 100 nm, and majority of the composition elements such as Fe, Cr, Ni, O and C were continuously changed towards the depth direction from the top surface. Characteristics of each layer show the operational history in the LHD and the mechanism of dusts formation by flaking of the layer. This work provides the completely new knowledge about the deposition layer in fusion devices.

Charge-exchange (CX) neutrals bombarding plasma-facing materials (PFMs) have large impacts not only on surface modifications of PFMs but also on plasma density control. To examine the effects of CX-neutrals on modifications of PFMs under specific plasma conditions, the controlled material probe experiments were performed in LHD. CX-neutrals incident parameters were evaluated quantitatively for each plasma condition from the viewpoint of microscopic damages. Microstructures of the samples exposed to LHD plasmas showed a significant dependence on the plasma density, working gas and sample temperature. These results are consistent with laboratory irradiation experiments and a modeling analysis. In addition, the suitability of this method as a CX-neutral monitor is discussed.

Detached plasmas can increase the radiated power fraction and thereby reduce the heat load on divertor tiles, which is important for steady-state fusion reactor operation. In LHD, divertor detachment is achieved through several means including gas puffing and the application of an m/n = 1/1 magnetic island using external perturbation coils. A recently upgraded imaging bolometer with a tangential view provides unique images of the plasma radiation, which give essential information about the source, three-dimensional (3-D) structure and dynamics of the radiation belt in the edge region. When compared to the results of edge modelling (EMC3/EIRENE) they can verify the model assumptions and identify the radial location of the radiation region, which is important because it impacts the compatibility between heat load and core plasma confinement and is dependent on impurity transport and magnetic topological effects on energy transport which are included in the 3-D edge model.

Chemical erosion of carbon materials and dust formation in low-temperature and neutral particle-dominated plasmas were investigated using high-pressure inductively coupled plasmas. Experiments were performed with Ar/H2/N2 mixture plasma irradiation to graphite targets. The addition of just a few percent of nitrogen gas to hydrogen led to significant suppression of carbon dust formation on the graphite target. From optical emission spectroscopy, CN band spectra were observed strongly in Ar/H2/N2 plasmas with a decrease of CH and C2 band emission intensity. These results showed that CN bond formation, which caused chemical erosion of carbon by producing volatile CN, HCN, and C2N2 particles, might have been a key suppression mechanism of the carbon particle aggregation.

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.

A lithium (Li) vapor injector for boundary control has been developed. A diverter covered with lithium is expected to reduce particle recycling. Recycling reduction is considered to be one of the triggers for the L–H transition. In this paper, the method of lithium dispersion is investigated under the assumption that the experiment is carried out in the Large Helical Device in National Institute for fusion Science, Japan (LHD). A performance test is performed on a prototype of the vapor injector. The amount of injected lithium was approximately 1% of the value expected from the vapor pressure data, due to the generation of lithium oxide. It is also found that nozzle temperature is quite important to suppress the Li dispersion.

Neutral particle behavior in the Large Helical Device heliotron has been investigated to conduct the effective particle control using the intrinsic helical divertor. It was revealed that the torus in–out asymmetry observed in the neutral pressure distribution depended on the divertor particle flux distribution, and thus, on the operational magnetic configuration. It was also revealed that the neutral pressure in the vacuum vessel in LHD was below 0.1 Pa. Degradation of the plasma confinement with increasing of the neutral pressure was observed, and that suggested the effective particle control is necessary for the sustaining of long discharge with high performance plasma and the further improvement of the confinement. The modification of the open helical divertor to the closed one was investigated for the particle control using helical divertor by using EIRENE code. Results of the calculation showed that proper rearrangement of divertor plates and additional components, such as dome structure make the neutral particles to be compressed well in the divertor region, and effective divertor pumping to be possible. Based on the simulation and experimental results, design of the closed helical divertor was completed and it will be partially installed in the Large Helical Device before the experimental campaign in 2010.

Plasma discharge operation with lithium coating suggests that the lithium effectively control neutral particles in the plasma periphery, which can lead to improvement of plasma parameters. The effect of lithium coating on the large helical device (LHD) for a closed helical divertor configuration is discussed from viewpoints of neutral particle and impurity ion transport in the plasma periphery. It shows that the closed helical divertor configuration can enhance the neutral particle density in the divertor region, which is enough to achieve efficient particle control, and that it can effectively confine neutral lithium atoms near divertor plates. A one-dimensional impurity (lithium) ion transport analysis along magnetic field lines on divertor legs indicates that the friction force due to the plasma flow from the main plasma is dominant over the thermal force caused by the temperature gradient on the divertor legs, which prevents lithium ion contamination in the main plasma and excessive cooling of the plasma temperature in an ergodic layer. The analysis shows that the lithium coating is compatible with LHD plasma discharge operation for the closed helical divertor configuration.

Magnetic flux surface measurements have been carried out in the Large Helical Device (LHD) in the standard magnetic field configuration with toroidal magnetic field strength up to 2.75 T. An electr...

Wall conditioning in the Large Helical Device (LHD) has been conducted successively since the first experimental campaign in 1998. The effects of wall conditioning on the vacuum condition, the plas...

A global particle balance study has been investigated in the Large Helical Device (LHD) in which the first wall and the divertor tiles are made of stainless steel (SUS-316L) and carbon, respectivel...

Easy access to the high-density regime without fatal disruptive phenomena is one of the important characteristics of the Large Helical Device (LHD). The operational density is considerably higher t...

The Large Helical Device (LHD) has been equipped with movable- and fixed-type material probe systems. Characterization studies of surface modifications on plasma-facing components (PFCs) ha...

To achieve an improvement of plasma confinement by an effective edge plasma control, the local island divertor (LID) was originally proposed in the National Institute for Fusion Science in the earl...

An interesting high-density operational regime with an internal diffusion barrier (IDB) has been extended to the helical divertor configuration in the Large Helical Device. The IDB is characterized...

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coup...

The Large Helical Device (LHD) is the world’s largest heliotron-type device with l = 2, m = 10 continuous superconducting helical coils and three pairs of superconducting poloidal coils. The major ...