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Impurity deposition and gas retention were investigated for material probes which were installed along toroidal and poloidal directions on plasma facing walls during the 4th experimental campaign in the large helical device (LHD). The carbon deposition was observed to be dominant on the surface of probes located far from the plasma. The retained amount of helium was also large in the probes located near the plasma, although that of hydrogen was large in the probe far from the plasma. The retained amounts of discharge gases and impurity gases were large near the anodes used for glow discharges, since the current density near the anode was large.

The erosion and deposition patterns on tungsten and tantalum test limiters exposed to the TEXTOR deuterium plasma containing a small amount of C impurity are simulated with the modified EDDY code. At the very top of the W and Ta limiters, there occurs neither erosion nor deposition, but the erosion proceeds slowly along the surface. When approaching the edge, the surface is covered by a thick C layer, which shows a very sharp boundary similar to the observation in surface measurements. In the erosion zone, the re-deposited carbon forms a W (Ta)–C mixed layer with small C concentration. Assumptions for chemical erosion yields of � 0.01 for W and

There are currently no universally accepted indications and criteria for additional surgical resection of the colorectum after endoscopic resection of the submucosal invasive cancer. The purpose of the present study is to establish accurate indications and criteria for additional surgical resection of the colorectum, based on the prediction of lymph node metastasis, after endoscopic resection of the submucosal invasive cancer. We investigated 140 submucosal invasive colorectal cancers and analyzed the pathologic factors of lymph node metastasis. The tumors were evaluated for pathologic factors in the invasive area of the submucosal carcinoma and were compared between the cases with lymph node metastasis and those without lymph node metastasis. Lymph node metastasis was observed in 13 cases (9%). Univariate logistic regression analysis showed that the depth of invasion, cribriform-type structural atypia, absence of lymphoid infiltration, lymphatic permeation, and venous permeation were statistically significant as risk factors for lymph node metastasis. Multivariate logistic regression analysis showed that the important risk factors included, in decreasing order, lymphatic permeation, absence of lymphoid infiltration, cribriform-type structural atypia, venous permeation, and depth of invasion. Submucosal invasion of 2 mm or more, and/or, depth of lymphatic permeation of 2 mm or more are risk factors for lymph node metastasis. The pathologic criteria based on our results for additional colectomy enables greater accuracy selection of patients who will undergo further surgical treatment after endoscopic resection.

Characteristics of transport in electron internal transport barriers (ITB) and in the vicinity of a rational surface with a magnetic island are studied with transient transport analysis as well as with steady state transport analysis. Associated with the transition of the radial electric field from a small negative value (ion-root) to a large positive value (electron-root), an electron ITB appears in the Large Helical Device [M. Fujiwara et al., Nucl. Fusion 41, 1355 (2001)], when the heating power of the electron cyclotron heating exceeds a power threshold. Transport analysis shows that both the standard electron thermal diffusivity, chie, and the incremental electron thermal diffusivity, chieinc (the derivative of normalized heat flux to temperature gradient, equivalent to heat pulse chie), are reduced significantly (a factor 5 10) in the ITB. The chieinc is much lower than the chie by a factor of 3 just after the transition, while chieinc is comparable to or even higher than chie before the transition, which results in the improvement of electron transport with increasing power in the ITB, in contrast to its degradation outside the ITB. In other experiments without an ITB, a significant reduction (by one order of magnitude) of chieinc is observed at the O-point of the magnetic island produced near the plasma edge using error field coils. This observation gives significant insight into the mechanism of transport improvement near the rational surface and implies that the magnetic island serves as a poloidally asymmetric transport barrier. Therefore the radial heat flux near the rational surface is focused at the X-point region, and that may be the mechanism to induce an ITB near a rational surface.

The spin alignment and cross section have been measured in the 12 C+ 16 O inelastic scattering for the 12 C(g.s.)+ 16 O(3 − 1 , 6.13 MeV) and 12 C(2 + 1 , 4.44 MeV)+ 16 O(g.s.) channels at E c.m. =17–23 MeV. In both channels, three weakly excited resonant structures are newly observed. Our result of spin alignment in the former channel well reproduces that of the previous study [Phys. Lett. B 120 (1983) 314]. In the latter channel the spin alignment is more or less enhanced at all intermediate structures. The spin alignment data integrated from forward to backward angles in this channel may indicate the presence of gross structure resonances.

Material probes have been installed at the inner walls along the poloidal direction in large helical device (LHD) from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to study the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was thoroughly cleaned by glow discharge conditioning and the number of main discharge shots increased. For the 3rd and 4th campaigns, graphite tiles were installed over the entire divertor strike region, and then the wall condition was significantly changed compared with the case of a stainless steel (SS) wall. It was seen that graphite tiles in the divertor were eroded mainly during main discharges, and the SS first wall mainly during glow discharges. During main discharges the eroded carbon was deposited on the entire wall. A reduction of metal impurities in the plasma was observed, which corresponds to the carbonized wall. The deposition thickness was great at the wall far from the plasma. Since the entire wall was carbonized, the amount of discharge gases retained such as H and He became large. In particular, helium retention was large at a position close to the anodes used for helium glow discharge cleanings. One characteristic of the LHD wall is a large retention of helium since the wall temperature is limited to below 368 K. In order to reduce the recycling of the discharge gas, wall heating is necessary.

To study chemical behaviors of energetic deuterium implanted into boron coating deposited by boronization in fusion devices, two types of boron coating film deposited on silicon and IG-430U were prepared by Plasma Chemical Vapor Deposition (PCVD) technique. Boron polycrystal was used as the reference sample. The chemical behavior of deuterium was investigated by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal adsorption spectroscopy).The 1.0 keV D{sub 2}{sup +} ions were implanted into the samples and the deuterium desorption behavior was studied by TDS. The TDS spectra showed that there were two deuterium release peaks at around 550 and 750 K, which were attributed to the release from deuterium trapped by boron and carbon, respectively. It was also found that most of implanted deuterium was trapped in carbon trapping site compared with boron one.In XPS measurements, the chemical shift of B-1s towards positive side was observed in the film on IG-430U after D{sub 2}{sup +} ion implantation. However, no chemical shifts were found in the film on silicon and boron polycrystal. In highly concentrated boron materials, even if deuterium was implanted into the boron materials, the amount of B-D bond was too low to be measured by XPS. This suggests that deuteriummore » implanted into highly pure boron materials wasn't almost trapped, so that the retention of deuterium in the boron materials would be reduced, compared that in carbon materials.« less

Ion heating experiments have been carried out in the large helical device using ECH (82.5, 84.0, 168 GHz, ≤1 MW), ICRF (38.5 MHz, ≤2.7 MW) and NBI (H° beam: 160 keV, ≤9 MW). The central ion temperature has been observed from the Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines, which are measured using a newly installed crystal spectrometer with a charge-coupled device. Recently, in ECH discharges, on-axis heating became possible. As a result, a high Te(0) of 6–10 keV and a high ion temperature of 2.2 keV were obtained at ne = 0.6×1013 cm−3. A clear increment of Ti was also observed with the enhancement of the electron–ion energy flow when the ECH pulse was added to the NBI discharge. These results demonstrate the feasibility towards ECH ignition. A clear Ti increment was observed also in ICRF discharges at low density ranges of (0.4–0.6)×1013 cm−3 with appearance of a new operational range of Ti(0) = 2.8 keV > Te(0) = 1.9 keV. In low power ICRF heating (1 MW), the fraction of bulk ion heating is estimated to be 60% of the total ICRF input power, which means Pi>Pe. Higher Ti(0), up to 3.5 keV, was obtained for a combined heating of NBI ( Te(0), whereas the Ti(0) remained at relatively low values of 2 keV in H2 and He NBI discharges due to less Pi. The main reasons for the high Ti achievement in the Ne discharges are: (1) 30% increment of deposition power, (2) increase in Pi/ni (five times, Pi/niPe/ne, Pi

Tungsten-coated carbon and copper was prepared by vacuum plasma spraying (VPS) and inert gas plasma spraying (IPS), respectively. W/CFC (Tungsten/Carbon Fiber-Enhanced material) coating has a diffusion barrier that consists of W and Re multi-layers pre-deposited by physical vapor deposition on carbon fiber-enhanced materials, while W/Cu coating has a graded transition interface. Different grain growth processes of tungsten coatings under stable and transient heat loads were observed, their experimental results indicated that the recrystallizing temperature of VPS-W coating was about 1400 °C and a recrystallized columnar layer of about 30 μm thickness was formed by cyclic heat loads of 4 ms pulse duration. Erosion and modifications of W/CFC and W/Cu coatings under high heat load, such as microstructure changes of interface, surface plastic deformations and cracks, were investigated, and the erosion mechanism (erosion products) of these two kinds of tungsten coatings under high heat flux was also studied.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.

Internal transport barriers with respect to electron thermal transport (eITB) were observed in the large helical device, when the electron cyclotron resonance heating (ECH) power was highly localized on the centre of a plasma sustained by neutral beam injection. The eITB is characterized by a high central electron temperature of 6–8 keV with an extremely steep gradient, as high as 55 keV m−1 and a low electron thermal diffusivity within a normalized average radius ρ≈0.3 as well as by the existence of clear thresholds for the ECH power and plasma collisionality.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

An internal transport barrier (ITB) was observed in the electron temperature profile in the Large Helical Device [O. Motojima et al., Phys. Plasmas 6, 1843 (1999)] with a centrally focused intense electron cyclotron resonance microwave heating. Inside the ITB the core electron transport was improved, and a high electron temperature, exceeding 10 keV in a low density, was achieved in a collisionless regime. The formation of the electron-ITB is correlated with the neoclassical electron root with a strong radial electric field determined by the neoclassical ambipolar flux. The direction of the tangentially injected beam-driven current has an influence on the electron-ITB formation. For the counter-injected target plasma, a steeper temperature gradient, than that for the co-injected one, was observed. As for the ion temperature, high-power NBI (neutral beam injection) heating of 9 MW has realized a central ion temperature of 5 keV with neon injection. By introducing neon gas, the NBI absorption power was increased in low-density plasmas and the direct ion heating power was much enhanced with a reduced number of ions, compared with hydrogen plasmas.

The superconducting machine LHD has conducted long pulse experiments for four years to achieve long-duration plasmas with high performance. The operational regime was largely extended in discharge duration and plasma density. In this paper, the plasma characteristics, in particular, plasma performance and impurity behaviour in long pulse discharges are described. Confinement studies show that global energy confinement times are comparable to those in short pulse discharges. Long sustainment of high performance plasma, which is equivalent to the previous achievement in other devices, was demonstrated. Long pulse discharges enabled us to investigate impurity behaviour in a long timescale. Intrinsic metallic impurity accumulation was observed in a narrow density window (2–3×1019 m−3) only for hydrogen discharges. Impurity transport study by using active impurity pellet injection shows a long impurity confinement time and an inward convection in the impurity accumulation window, which is consistent with the intrinsic impurity behaviour. The pulsed neon gas injection experiment shows that the neon penetration into the plasma core is caused by the inward convection due to radial electric field. Finally, impurity accumulation control with an externally induced magnetic island at the plasma edge was demonstrated.

Divertor footprints have been identified within a few mm in accuracy after 10 000 shots. This is the large merit of the large helical device divertor for erosion/deposition studies due to high reproducibility with an external superconducting coils system. Helical distribution of divertor erosion is compared with the prediction from magnetic field characteristics. The measured net erosion depth is found to be about a factor 3 less than the estimated one. Numerical simulations have revealed the net erosion to be very sensitive to deposition of C impurity in the plasma. Eroded carbon atoms are mainly redeposited near the divertor tiles, and partly deposited near the divertor strike point, forming a mixed layer with promptly deposited metals. Deposited metals accumulate locally at the edge of microscale open pores and around grains of graphite. This kind of metals sink possibly plays an important role as an impurity source after the tiles installation. This aspect of ‘microscopic-PSI study’ is very informative for understanding macroscopic-PSI.

Surface modification and helium retention on powder metallurgy tungsten and plasma sprayed tungsten coated CFC exposed by low energy (100 eV) high flux (3.83×10 21 –1.20×10 22 m −2 s −1 ) helium plasma have been investigated. Fluence is about 10 26 He m −2 . It is found that fine surface morphology change occurs by the helium exposure. There is little difference of the surface modification between the powder metallurgy tungsten and the plasma sprayed tungsten. In the case that the surface morphology do not changed, two helium desorption peak appeared. The amount of desorption of helium is larger than that of deuterium from tungsten under the same exposure condition by heating up to 1273 K by TDS. In the case that fine surface morphology change occurs, desorption peak at high temperature newly appears. Surface modification such as blister is not formed on the surface exposed to helium under the same exposure condition which the blister is formed by deuterium exposure.

From viewpoints of thermal shock resistance, control of plasma particles, low activation, thermal efficiency and so on, it is planned to use heat-resisting metals and ceramic composites as plasma facing materials for the next experimental, demonstrative and commercial fusion reactors. In this study, a tungsten material and SiC/SiC composites were joined with oxygen free copper as a heat sink material using foils of titanium and copper, and a molybdenum plate was also inserted for the relaxation of residual thermal stresses in the case of a SiC/SiC-copper joint. The divertor model specimens using the joining materials were manufactured and heat load tests were carried out. Thermal cracks in the tungsten material and delaminating cracks at the joining boundary of SiC/SiC composites were observed during several heat load tests. Therefore, tungsten and SiC/SiC composites need to be improved further with respect to the thermal shock resistance, thermal conductivity and fracture toughness.

By changing the distribution of the coil currents in the helical coils of the large helical device (LHD) the plasma could be scraped off on the stainless steel inboard wall at the last closed flux surface, eliminating the ergodic edge region and helical divertor (HD) plasmas. During such NBI-heated discharges slow oscillations (1–2.5 Hz) have been observed in the major global parameters. These oscillations exhibit many similarities to the ‘breathing’ relaxation phenomenon observed during long pulse experiments in LHD when the divertor material was stainless steel. These similarities include a large core radiation fraction, significant radiation from the metallic impurities and dependence of the oscillation frequency on density. In contrast, discharges using the graphite HD show hollow radiation profiles and reduced levels of radiation from heavy impurities. Modeling shows that differences in the density dependence of the oscillation frequency between the wall limiter and HD cases can be attributed to differences in the impurity diffusion between the core and ergodic edge regions.

To reveal interaction between boron coating film and energetic hydrogen isotopes, chemical states of deuterium implanted into the boron coating films deposited on graphite by using DC glow discharge in a diborane diluted helium have been studied. The XPS measurements showed the possibility of the formation of B–D and C–D bonds by D2+ ion implantation. The TDS experiments indicated that deuterium desorption took place in two stages, which were attributed to the desorption processes from the trapping sites of B–D and C–D bonds. The results of the XPS measurements after D2+ ion implantation revealed that the peak areas of the B–D and C–D bonds began to decrease at 310 and 473 K and the peaks almost disappeared at 873 and 1073 K, respectively. These results suggest that the trapping site of the B–D bond is largely influenced by thermal annealing effects compared with that of the C–D bond.

Selection of candidate materials for plasma facing material (PFM) in HT-7U device and plasma–wall interactions are critically important to reach high plasma performance. Based on concentrated research on multi-element doped graphite containing B, Si and Ti, two kinds of doped graphites have been chosen as candidates for PFM in HT-7U. Doped graphite GBST1308 with the dopant concentration of 1% B, 2.5% Si, 7.5% Ti was developed as low-Z PFM for reducing the chemical sputtering and suppressing the radiation enhanced sublimation, and successfully used as the new limiter material in last two campaigns of HT-7 tokamak experiments. Doped graphite with the composition of 2.5% Si, 7.5% Ti has improved mechanical properties and thermal conductivity of 314 W/m K at room temperature. TDS and high heat flux experiments results demonstrated that such doped graphites are promising candidate plasma facing components for HT-7U. 2003 Elsevier Science B.V. All rights reserved.

To investigate the effect due to localized high heat flux pulses onto plasma facing components, light pulses from a ruby laser (15 J in maximum energy, 0.3 ms in duration) were injected onto test-limiter (C, W, and Ta) surfaces immersed in TEXTOR edge plasmas. When the laser pulses irradiated the C limiter surface, the CVI signal from the plasma core increased slightly. In laser injection onto high-Z limiters, the increment of the quasi-continuum around 5 nm of the Ta signal was larger than that of the W signal. With the laser focusing down to 2 mm in diameter, small high-Z particles were found to be ejected from the limiter surfaces which penetrated into the plasma beyond the LCFS. However, the plasma withstood the core cooling due to the radiation loss of high-Z impurity. The amounts of impurities in the core were also estimated by a laser blow-off experiment.

For the third experimental campaign of LHD, graphite tiles were installed in entire region of divertor strike region. Several material probes of SS 316L and graphite were placed along the poloidal direction at #7 toroidal sector. These probes were extracted after the campaign, and impurity deposition on the probe and retention of discharge and impurity gases were analyzed by AES and TDS, respectively. Major impurity species deposited on the probe were C, Fe and O. The iron deposition rate per one main discharge was approximately a half of that in the second campaign. The carbon deposition was significantly large, and the carbon concentration was as high as 90 at.%. These results correspond to the reduction of metal impurity level in the plasma. The retention of discharge gas and impurities doubled compared to the case of the second campaign, due to the carbonized wall. Significant retention of helium was observed, which is one of the characteristics of the LHD wall. The retention of helium is due to charge exchanged He during He main discharge shots and implantation of He ion during He glow discharges.

A divertor head of local island divertor (LID) that utilizes an m/n=1/1 island was installed on the large helical device (LHD) after the fifth experimental campaign in 2001–2002. Up to the fifth campaign, the effect of the LID magnetic configuration on plasma performance was studied without the divertor head. It was found that the island itself deteriorates plasma parameters such as electron temperature, stored energy, etc., and that the ergodic layer around the natural helical separatrix plays a role on preventing impurities from penetrating into the core plasma.

β-Ray-induced X-ray spectrometry was applied to determine the amounts and depth profiles of tritium captured by graphite tiles used in the PTE1 campaign and by carbon fiber composite tiles in the DTE1 campaign of JET. Although the X-ray spectra observed for both carbon tiles consisted of characteristic and bremsstrahlung X-rays, the intensities and shapes of the X-ray spectra were quite different between these tiles. The analyses of the X-ray spectra showed that the amount of tritium retained in the latter tile was on average 15 500 times greater than that in the former tile, and the tritium depth profile extended to 60 μm for the latter tile compared to 5 μm for the former tile. In addition, characteristic X-rays corresponding to Cl, Cr, Fe, Ni and Mo appeared in the latter tile, where Ni showed the highest concentration, indicating that considerable erosion and deposition of vacuum vessel materials took place in the DTE1 campaign.

We have performed a Monte Carlo simulation of transport of atomic and molecular hydrogen and hydrocarbons released from W,Ta and C in edge plasmas of TEXTOR. The simulation derives thermal,dissociation and reflection components from the observed radial distribution of Dc line intensity around W–Ta and W–C twin test limiters. The Dc intensity in front of the limiter is dominated by thermal reemission of atomic hydrogen,whereas at the position well away from the limiter,dissociation of reemitted molecular hydrogen and reflection of incident hydrogen are important. The simulation allows us to estimate the surface temperature and material dependence of atomic and molecular fractions of hydrogen reemitted from the limiter. For Ta,hydrogen is reemitted as a mixture of atoms and molecules at 1200 K and the atomic fraction increases with increasing temperature (1400 K),whereas for W,the atomic release dominates the reemission process at the temperatures. 2003 Elsevier Science B.V. All rights reserved.

Plasma sprayed tungsten coated carbon fiber composites (VPS-W/CX-2002U) and powder metallurgy tungsten (PM-W) have been exposed to a high deuterium flux (≃1022 m−2 s−1) with low energy (100 eV) in a range from 708 to 843 K. Surface modification and deuterium retention after the exposure have been investigated to prove the suitability of such materials in fusion devices. Blisters are formed on the PM-W by deuterium irradiation with a fluence of 7.5×1025 m−2. The amount of blisters and their average size increase with an increase of a fluence to 3.00×1026 m−2. On the other hand, no modification is observed on VPS-W/CX-2002U. Desorption of VPS-W/CX-2002U irradiated by deuterium is different from that of PM-W. The peak temperature of D2 and HD release from PM-W is about 703 K. However, the desorption curve of VPS-W/CX-2002U gradually increases with increasing temperature up to 1273 K.

The high-Z impurity concentration of tungsten (W) and tantalum (Ta) released from a twin-test-limiter half made of W and the other half made of Ta was investigated by inserting the twin-test-limiter into the edge plasmas of the TEXTOR-94 tokamak. To study the spectra of W and Ta in the extreme ultraviolet (XUV) range, these high-Z materials were introduced into the plasma by a laser blow-off technique. The quasi-continuum spectrum near 5 nm by pure Ta shifted about 0.2 nm to the longer wavelength side than that by pure W. The intensities were about the same for both material for the identical discharge parameters. The increment in temperature for the Ta limiter facing the ion drift side was about 1.4 times higher than that of W for identical plasma exposure conditions. The larger increase in temperature can be quantitatively explained by a smaller thermal conductivity of Ta. In the test-limiter experiment, the position of the peak in the XUV spectrum was unaffected by the orientation of the limiter, which corresponded to an equal influence of the particle emissions from both sides of the limiter.

Intrinsic impurity behaviour in neutral beam heated LHD long pulse discharges with a discharge duration of 10 s was investigated. Spectroscopic and bolometric measurements showed a remarkable temporal increase of core radiation due to metallic impurities. Central impurity accumulation was found in a narrow plasma density region (e = 1–3×1019 m−3) for both density ramp-up discharges and constant density discharges. In the density ramp-up discharges, high-Z impurities were accumulated in the low-density region and diffused out from the core plasma in the high-density region. The impurity behaviour was compared with neoclassical predictions.

Multi-element doped graphite, GBST1308 has been developed as a plasma facing material (PFM) for high heat flux components of the HT-7U device. The thermal performance of the material under steady-state (SS) high heat flux was evaluated under actively cooling conditions, the specimens were mechanically joined to copper heat sink with supercarbon sheet as a compliant layer between the interfaces. The experiments have been performed in a facility of ACT (actively cooling test stand) with a 100 kW electron gun in order to test the suitability and the loading limit of such materials. The surface temperature and bulk temperature distribution of the specimens were investigated. The experimental results are very encouraging that when heat flux is not more than 6 MW/m2, the surface temperature of GBST1308 is less than 1000 °C, which is the lowest, compared with IG-430U and even with CX-2002U (CFC); The primary results indicate that the mechanically-joined material system by such as a proper design as thin tile, super compliant layer, GBST as a PFM and copper-alloy heat sink, can be used as divertor plates for HT-7U in the first phase.

The utility of ICRF heating in the LHD was demonstrated in the third campaign carried out in 1999. This paper summarizes the investigations made in 2000 with a focus on the optimization of ICRF heating. The flexibility of the LHD magnetic configuration was fully utilized as a key factor in the investigations. The experiments include (a) scan of magnetic field intensity, (b) scan of aspect ratio, (c) scan of magnetic axis shift, (d) scan of quadrupole magnetic field, and (e) cancellation of magnetic island. The performance of the ICRF heating was thus optimized with respect to magnetic parameters, while optimization was achieved mainly with respect to heating regime and wall conditioning in the third campaign. Some of these parameters are directly related to the orbit of trapped particles. Therefore, the relation between particle orbits and the performance of ICRF heating is also addressed in this paper based on analyses of the high-energy tail of ions.

Divertor plasma characteristics in the Large Helical Device (LHD) have been investigated mainly by using Langmuir probes. The three-dimensional structure of the helical divertor, which is naturally produced in the heliotron-type magnetic configuration, is clearly seen in the measured particle and power deposition profiles on the divertor plates. These observations are consistent with the numerical results of field line tracing. The particle flux to the divertor plates increases almost linearly with the line averaged density. The high-recycling regime and divertor detachment, which are observed in tokamaks, have not been observed even during high density discharges with low input power. Both electron density and temperature decrease with increasing radius in the stochastic layer with open field lines, and at the divertor plate they become fairly low compared with those at the last closed flux surface. This means the reduction of pressure along the magnetic field lines occurs in the open field line region in LHD.

In the Large Helical Device (LHD), active edge plasma control using two types of divertor, helical divertor (HD) and the local island divertor (LID), is planned, respectively. The former is intrinsic in the heliotron-type magnetic configuration. In the latter case, large pumping efficiency is expected by making particle recycling localized toroidally using a divertor head and an externally produced m = 1, n = 1 island. In this paper, the divertor properties in the HD configuration are reviewed. The numerical result for design of LID is described. The results of the LID experiment performed in the Compact Helical System are also briefly reviewed.

The experimentally optimized magnetic field configuration of the Large Helical Device [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)], where the magnetic axis is shifted inward by 15 cm from the early theoretical prediction, reveals 50% better global energy confinement than the prediction of the scaling law. This configuration has been investigated further from the viewpoints of high energy particle confinement and magnetohydrodynamic (MHD) stability. The confinement of high energy ions is improved as expected. The minority heating of ion cyclotron range of frequency was successful and the heating efficiency was improved by the inward shift. The confinement of passing particles by neutral beam injection was also improved under low magnetic field strength, and there could be obtained an almost steady high beta discharge up to 3% in volume average. This was a surprising result because the observed pressure gradient exceeded the Mercier unstable limit. The observed MHD activities became as high as beta but they did not grow enough to deteriorate the confinement of high energy ions or the performance of the bulk plasma, which was still 50% better than the scaling. According to these favorable results, better performance would be expected by increasing the heating power because the neoclassical transport can also be improved there.

The results of transport modeling for breathing oscillations found recently in the Large Helical Device (LHD) [1] are presented. In the model elaborated the time evolution of the central plasma temperature and density are determined from the energy and particle balance in the plasma core, the boundary temperature is computed from the edge heat balance and the transport of Fe ions over the minor radius is described in a 1-D approximation with a self-consistently computed source due to ionisation of neutrals sputtered at divertor plates. The results of computation substantiate the qualitative explanation proposed earlier [2] and allow to elaborate a physical picture of breathing oscillations consistent with experimental observations. The calculations reproduce well the experimental time evolution of different parameters, their phasing and the variation of the amplitude and period of oscillations with plasma characteristics.

Macroautophagy (autophagy) is a bulk protein-degradation system ubiquitously conserved in eukaryotic cells. During autophagy, cytoplasmic components are enclosed in a membrane compartment, called an autophagosome. The autophagosome fuses with the vacuole/lysosome and is degraded together with its cargo. Because autophagy is important for the maintenance of cellular homeostasis by degrading unwanted proteins and organelles, identification of autophagosome cargo proteins (i.e., the targets of autophagy) will aid in understanding the physiological roles of autophagy. In this study, we developed a method for monitoring intact autophagosomes ex vivo by detecting the fluorescence of GFP-fused aminopeptidase I, the best-characterized selective cargo of autophagosomes in Saccharomyces cerevisiae. This method facilitated optimization of a biochemical procedure to fractionate autophagosomes. A combination of LC-MS/MS with subsequent statistical analyses revealed a list of autophagosome cargo proteins; some of these are selectively enclosed in autophagosomes and delivered to the vacuole in an Atg11-independent manner. The methods we describe will be useful for analyzing the mechanisms and physiological significance of Atg11-independent selective autophagy.

Atg5 is covalently modified with Atg12 via reactions that are similar to ubiquitination, and it noncovalently interacts with Atg16. Formation of the Atg12–Atg5–Atg16 complex is essential for its E3-like function: facilitation of Atg8 transfer from Atg3 to phosphatidylethanolamine at autophagic membranes. Structural studies on the Atg12–Atg5–Atg16 complex revealed that the unique architecture of this protein complex is totally distinct from the other E3 enzymes. The Atg12–Atg5–Atg16 complex interacts directly with Atg3 via Atg12, and enhances the conjugase activity of Atg3 by rearranging its catalytic center, while it is targeted to the membranes via Atg5 and Atg16, and promotes the transfer of Atg8 from Atg3 to the membranes.

Autophagy mediates the selective degradation of various targets, such as aggregated proteins, damaged or superfluous organelles, and invading microbes. Specific adaptors, also known as receptors, play a critical role in target recognition in autophagy. Almost all adaptors possess at least one Atg8 family interacting motif (AIM) that interacts directly with Atg8 family proteins, as a prerequisite for selective packaging of targets into autophagosomes. Recent studies have revealed not only the divergence of AIM sequences and their interaction with Atg8 family proteins, but also the regulation of the AIM–Atg8 interaction through phosphorylation.

This paper reports on the behaviour of high-energy ions created by ion cyclotron range of frequency (ICRF) heating on the Large Helical Device (LHD). In the third experimental campaign conducted in 1999, it was found that minority heating has good heating performance, and high-energy particles were observed. In the fourth campaign in 2000, the temporal behaviour of high-energy ions was investigated in the minority heating regime using turnoff or modulation of ICRF power. The time evolution of the high-energy particle distribution was measured using a natural diamond detector (NDD) and a time-of-flight neutral particle analyser (TOF-NPA). It was found that the count number of higher-energy particles declines faster than that of lower-energy particles after ICRF turnoff. In the modulation experiments, the phase difference of the flux of high-energy particles with respect to the ICRF power modulation increased with energy. These results were explained qualitatively by the Fokker-Planck equation with a simple model. The pitch-angle dependence of the distribution function was also measured in the experiment by changing the line of sight of the TOF-NPA, and an anisotropy of the high-energy tail was found. This anisotropy was reproduced by solving the bounce-averaged Fokker-Planck equation. The second harmonic heating was conducted successfully for the first time in the LHD in high-β plasma, and high-energy particles were also detected in this heating regime.

Thermal performances of an advanced mechanically joined module (MJM) under steady state high heat fluxes of 2.0–4.25 MW/m2 and thermo-mechanical analysis are described. The advanced MJM was designed to apply for a helical divertor plate of a large helical device (LHD) at the next experimental phase II. The advanced MJM has a unified armor/heat sink made of carbon/carbon composite different from the normal MJM with a separated copper heat sink. To evaluate the thermal performance of the advanced MJM, short pulse high heat flux test up to 5.4 MW/m2 and steady state high heat flux tests up to 4.25 MW/m2 have been carried out using a test facility ACT. Moreover, the thermal fatigue test of the advanced MJM up to 150 cycles under steady state high heat flux of 2.5 MW/m2 has been performed. After theses tests, no apparent damage and no cracking on the armor tile were observed although there was a little increase in the armor/heat sink temperature during the thermal fatigue test. To evaluate the thermal stress in the armor/heat sink of the MJM during steady state high heat flux test and to optimize the structure, thermo-mechanical analyses are done using a 3D CAD.

During the first two years of the LHD experiment the following results have been achieved: (i) higher Te (Te(0) = 4.4 keV at ne = 5.3 × 1018 m-3 and Pabs = 1.8 MW); (ii) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ne = 6.5 × 1019 m-3 and Pabs = 2.0 MW); (iii) higher stored energy, Wpdia = 880 kJ at B = 2.75 T. High performance plasmas have been realized in the inward shifted magnetic axis configuration (R = 3.6 m) where helical symmetry is recovered and the particle orbit properties are improved by a trade-off of MHD stability properties due to the appearance of a magnetic hill. Energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 by up to a factor of 1.6 and was comparable with the ELMy H mode confinement capability in tokamaks. This confinement improvement is attributed to configuration control (inward shift of the magnetic axis) and to the formation of a high edge temperature. The average beta value achieved reached 2.4% at B = 1.3 T, the highest beta value ever obtained in a helical device, and so far no degradation of confinement by MHD phenomena has been observed. The inward shifted configuration has also led to successful ICRF minority ion heating. ICRF powers up to 1.3 MW were reliably injected into the plasma without significant impurity contamination, and a plasma with a stored energy of 200 kJ was sustained for 5 s by ICRF alone. As another important result, long pulse discharges of more than 1 min were successfully achieved separately with an NBI heating of 0.5 MW and with an ICRF heating of 0.85 MW.

Radio frequency (rf) siliconization has been carried out on the HT-7 superconducting tokamak in the presence of a high magnetic field, which is a try on superconducting tokamaks. Three different procedures of rf siliconization have been tested and a very promising method to produce high quality silicon films was found after comparing the film properties and plasma performance produced by these three different procedures. The Si/C films are amorphous, semitransparent, and homogeneous throughout the layer and adhere firmly to all the substrates. The advantages of silicon atoms as a powerful radiator and a good oxygen getter have been proved. An outstanding merit of rf siliconization to superconducting devices is its fast recovery after a serious degradation of the condition due to the leakage of air to good wall conditions. A wider stable operation region has been obtained and plasma performance is improved immediately after each siliconization due to significant reduction of impurities. Energy confinement ...

A power modulation experiment was conducted in the third campaign of the LHD. In a conventional analysis of the modulation experiments, the energy confinement time and heating efficiency are taken as constant, disregarding their dependence on the plasma parameters. In this paper, their dependence on the plasma temperature and heating power is taken into consideration to improve the analysis of the power modulation experiments. Several models, with differing dependence on the plasma parameters, have been examined. There have been several reports suggesting that the transport coefficients obtained from the dynamical method are different from those obtained from power balance analyses. This paper finally concludes that the energy confinement time obtained from the power modulation experiments well agrees with that obtained from the power balance analysis, made including temperature dependence both in energy confinement time and heating efficiency.

The ICRF heating experiments conducted in 1999 in the third experimental campaign on LHD are reported, with an emphasis on the optimization of the heating regime. Specifically, an exhaustive study of seven different heating regimes was carried out by changing the radiofrequency relative to the magnetic field intensity, and the dependence of the heating efficiency on H minority concentration was investigated. It was found in the experiment that both ion and electron heating are attainable with the same experimental set-up by properly choosing the frequency relative to the magnetic field intensity. In the cases of both electron heating and ion heating, the power absorption efficiency depends on the minority ion concentration. An optimum minority concentration exists in the ion heating case while, in the electron heating case, the efficiency increases with concentration monotonically. A simple model calculation is introduced to provide a heuristic understanding of these experimental results. Among the heating regimes examined in this experiment, one of the ion heating regimes was finally chosen as the optimized heating regime and various high performance discharges were realized with it.

The energy confinement and thermal transport characteristics of net current free plasmas in regimes with much smaller gyroradii and collisionality than previously studied have been investigated in the Large Helical Device (LHD). The inward shifted configuration, which is superior from the point of view of neoclassical transport theory, has revealed a systematic confinement improvement over the standard configuration. Energy confinement times are improved over the International Stellarator Scaling 95 by a factor of 1.6 ±0.2 for an inward shifted configuration. This enhancement is primarily due to the broad temperature profile with a high edge value. A simple dimensional analysis involving LHD and other medium sized heliotrons yields a strongly gyro-Bohm dependence (τEΩ ∝ ρ*-3.8) of energy confinement times. It should be noted that this result is attributed to a comprehensive treatment of LHD for systematic confinement enhancement and that the medium sized heliotrons have narrow temperature profiles. The core stored energy still indicates a dependence of τEΩ ∝ ρ*-2.6 when data only from LHD are processed. The local heat transport analysis of discharges dimensionally similar except for ρ* suggests that the heat conduction coefficient lies between Bohm and gyro-Bohm in the core and changes towards strong gyro-Bohm in the peripheral region. Since the inward shifted configuration has a geometrical feature suppressing neoclassical transport, confinement improvement can be maintained in the collisionless regime where ripple transport is important. The stiffness of the pressure profile coincides with enhanced transport in the peaked density profile obtained by pellet injection.

Recent large helical device experiments revealed that the transition from ion root to electron root occurred for the first time in neutral-beam-heated discharges, where no nonthermal electrons exist. The measured values of the radial electric field were found to be in qualitative agreement with those estimated by neoclassical theory. A clear reduction of ion thermal diffusivity was observed after the mode transition from ion root to electron root as predicted by neoclassical theory when the neoclassical ion loss is more dominant than the anomalous ion loss.

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position R_ax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with R_ax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.

During experiments in LHD using stainless steel divertor plates, a slow (~1 s) cyclic oscillation in the plasma parameters known as `breathing' plasma was observed during NBI heated long pulse discharges. Using an average ion, corona equilibrium model for the iron impurity cooling rate, the iron impurity density profile is calculated for 0.0

Pellet injection has been used as a primary fuelling scheme in the Large Helical Device. With pellet injection, the operational region of NBI plasmas has been extended to higher densities while maintaining a favourable dependence of energy confinement on density, and several important values, such as plasma stored energy of 0.88?MJ, energy confinement time of 0.3?s, ? of 2.4% at 1.3?T and density of 1.1 ? 1020?m -3, have been achieved. These parameters cannot be attained by gas puffing. Ablation and the subsequent behaviour of the plasma have been investigated. The measured pellet penetration depth estimated on the basis of the duration of the H? emission is shallower than the depth predicted from the simple neutral gas shielding (NGS) model. It can be explained by the NGS model with inclusion of the effect of fast ions on the ablation. Just after ablation, the redistribution of the ablated pellet mass was observed on a short timescale (~400?ms). The redistribution causes shallow deposition and low fuelling efficiency.