Search

An ion cyclotron range of frequency (ICRF) heating experiment was conducted in the third campaign of LHD in 1999. 1.35 MW of ICRF power were injected into the plasma and 200 kJ of stored energy were obtained, which was maintained for 5 s by ICRF power only after the termination of ECH. The impurity problem was so completely overcome that the pulse length was easily extended to 68 s at a power level of 0.7 MW. The utility of a liquid stub tuner in steady state plasma heating was demonstrated in this discharge. The energy confinement time of the ICRF heated plasma has the same dependences on plasma parameters as those of the ISS95 stellarator scaling with a multiplication factor of 1.5, which is a high efficiency comparable to that of NBI. Such an improvement in performance was obtained by various means, including: (a) scanning of the magnetic field intensity and minority concentration, (b) improvement of particle orbits due to a shift of magnetic axis and (c) reduction of the number of impurity ions by means of titanium gettering and the use of carbon divertor plates. In the optimized heating regime, ion heating turned out to be the dominant heating mechanism, unlike in CHS and W7-AS. Owing to the high quality of the heating and the parameter range being extended far beyond that of previous experiments, the experiment can be regarded as the first complete demonstration of ICRF heating in stellarators.

The correlation between the local WI light emission from W limiters and the W quasi-continuum light emission from the core plasma measured by the XUV spectrometer was investigated in the TEXTOR-94 limiter tokamak. The W quasi-continuum light emission showed a linear dependence against the line spectrum emission from neutral W from the metallic test limiter immersed in neutral beam (NBI) heated plasmas. The surface of the poloidal limiter made of vacuum-plasma-spray coating of W (VPS-W) was covered by a carbon layer during the time it was kept in the shadow of graphite ALT-II limiter. As a result of the carbon coating, the contribution to the W quasi-continuum intensity by the poloidal limiter was smaller than that by the test limiter, despite the fact that the surface area of the poloidal limiter facing the plasma was about seven times larger than that of the test limiter.

The large helical device (LHD) is the largest heliotron type superconducting device. Its operation was started on 31 March 1998. Three experimental campaigns have been completed until the end of 1999. Wall conditioning mainly by cleaning discharges using ECRF or glow discharges worked well even without high temperature baking. The plasma production with ECRH and auxiliary heating with NBI and/or ICRF in the LHD configuration equipped with open helical divertor were well performed. The divertor material was SS316L in the first and second campaigns, and was replaced by the graphite in the third campaign. The influences of the different divertor materials were investigated. Our understanding of the edge and the divertor plasma has progressed. Long-pulse discharges 80 and 68 s heated by NBI (0.5 MW) or ICRF (0.9 MW) have been achieved, respectively. No severe limitation of the duration has appeared.

The first and second experimental campaigns in large helical device (LHD) were carried out from March to December 1998. Before the each campaign, the material probes were placed at the inner wall of vacuum vessel along the poloidal direction. After each campaign, change of surface morphology, impurity depth profile and gas desorption of the samples were examined. In the first experimental campaign, the ECR discharge cleanings were employed. After the first experimental campaign, the surface was modified by the deposition of sub-micron particles, and the concentrations such as oxygen and carbon were still high. Impurity gas desorption was also large in the sample at the port. In the second experimental campaign, the glow discharge cleanings were employed and the number of main discharge shots increased. After the second experimental campaign, no significant deposition took place except for the position close to the divertor leg, and the oxygen impurity level was reduced. In every sample, the helium was retained by the helium glow discharge. In addition, the amount of gas desorption was considerably reduced even in the sample at the port. In the second experimental campaign, the wall conditionings largely progressed by using the glow discharge cleanings and the increase of main discharge shots with a high heating power.

In TEXTOR-94, experiments have been performed with the upper and lower poloidal limiter blocks made of vapour sprayed (VSP) tungsten (about 0.5 mm) deposited on graphite with a rhenium interlayer. A series of discharge conditions have been performed (density scan, scan of the auxiliary heating power, radius scan). There has been found no restriction for operation at any density with auxiliary heating. For Ohmic conditions the same density with testlimiters could be reached. Under siliconized conditions no severe accumulation of tungsten in the plasma centre could be detected. The blocks could in general stand surface temperatures below 1700 K. Most of them survived also temperatures above 3000 K without exfoliation. However, some blocks showed severe damage by melting or exfoliation probably due to insufficient contact of the tungsten layer with the graphite.

Applicability of a newly developed β-ray induced X-ray spectrometry (BIXS) has been examined to measure nondestructively tritium retained on/in the graphite samples. Examination was carried out by using the graphite plates irradiated with tritium ions and an ALT-II limiter tile exposed to D-plasmas in TEXTOR. For the former samples, a sharp intense peak and a broad weak peak appeared clearly in the spectra; the former peak was attributed to the characteristic X-rays from argon used as a working gas, and the latter peak was assigned to the bremsstrahlung X-rays from sub-surface layers of graphite. On the other hand, for the latter sample, a rather weak characteristic X-ray peak was observed along with a diminutive bremsstrahlung X-ray peak. Although the intensities of those X-rays differed from spot to spot, the tritium levels retained on the limiter tile were determined to be 58–132 Bq / cm 2 . It was concluded, therefore, that valuable information on the amount and the distribution of tritium retained on/in the wall materials can be nondestructively obtained by using the BIXS.

In the large helical device (LHD) radiation profiles measured using arrays of resistive metal, foil bolometers are used to investigate the change in radiation resulting from a replacement of stainless steel divertor plates with graphite tiles between the second and third experimental campaigns. In particular, for the magnetic configuration of R axis = 3.6 m and at line averaged densities below 3 x 10 19 m 3 , a reduction of the radiated power fraction from 35% in the second campaign to down to 12% in third campaign was observed. Comparing similar shots (in terms of discharge parameters) from the second and third cycles the core (0 < r/a < 0.79) radiation fraction was reduced from 44% to 30% of the total radiation. Spectroscopic measurements show a corresponding decrease in radiation from iron. Comparing long pulse discharges the radiated power fraction reduced from greater than 67% during the density limiting oscillation observed in the second campaign to 20% during the third campaign with a reduction in the respective core radiation fractions from 63% to 37%.

Operational density regime of long pulse discharges heated by neutral beam injection (NBI) alone on the large helical device (LHD) is greatly extended by the replacement of stainless steel (SS) divertor plates with graphite tiles resulting in a remarkable reduction of metal impurity radiation. The plasma density can be controlled by gas puffing up to 6×10 19 m −3 . Particle balance is analyzed by a global particle balance equation for helium and hydrogen discharges with high density. For helium discharges, the plasma particle inventory can be almost completely accounted for by the integrated gas puff input. A strong wall pumping is observed for hydrogen only discharges and it is found that the recycling coefficient at the start-up of the discharge is less than 0.6. The recycling coefficient increases with the wall particle inventory and approaches unity for high wall loading (23 Pa m 3 ) .

Mutual contamination between C and W, resulting from the simultaneous use of these materials as plasma facing components, is simulated by means of a computer simulation code, Erosion and Deposition based on Dynamic model (EDDY). W deposition on C rapidly increases the reflection coefficient for D and C impurity. In comparison between the calculation and a C-W twin test limiter experiment in TEXTOR-94, C release from the C side of the limiter is dominated by reflection of C impurity from the W deposits, in addition to physical sputtering of C; chemical erosion is strongly suppressed. Due to the dynamic effect which makes C-W mixed layer, C deposition on W gradually changes the reflection coefficient and sputter yields. Formation of a sharp boundary between erosion and C deposition zones on the W side of the limiter is well reproduced by simulation. Local redeposition patterns of C and W on the limiter surface are also calculated.

We have developed and constructed an ECH system for the large helical device (LHD). The ECH system consists of 0.5 MW, 84 GHz range and 168 GHz gyrotrons, high voltage power supplies, long distance transmission lines, and in-vessel quasi-optical antennas. It has been improved step by step. At the third campaign of LHD experiments, three 84 GHz range (two 82.6 GHz and one 84 GHz) and three 168 GHz gyrotrons are operated and ECH power can be injected from four antennas vertically and two horizontally. This complicated system is remotely controlled and monitored by fully GUI (Graphical User Interface) control panels realized on PC via TCP (transmission control protocol) communication. Over 10 000 shots of gyrotron power have been injected steadily into the LHD during the experimental campaigns on this system. One line of the system (84 GHz line) is specially prepared for the experiments of steady state plasma production. Using this line, plasma sustainment for 2 min was successfully achieved by only ECH power. Injected ECH power was 50 kW with 95% duty factor. The electron density and temperature of the sustained plasma are measured to be 0.3–0.5×1018 m−3 and ∼650 eV. Ion temperature measured by Doppler broadening of the impurity radiation line was kept constant at ∼300 eV during RF injection.

In order to investigate the impurity release and surface modification on a W/C twin test limiter, made of a half of W and the other half of C, exposed to the edge plasma of TEXTOR-94, simulation calculations of ion-surface interaction are conducted by a Monte Carlo code. According to the calculations, experimentally observed spatial distributions of WI and CII line intensities around the W side of the limiter can be explained by physical sputtering of W, reflection of bombarding C ions and physical sputtering of implanted C. The CII line emission, resulting from thermal C atoms, around the C side of the limiter is suppressed by deposition of W, and the reflection of C ions from W deposited on C causes the CII intensity to decay more slowly than that from C without the deposition. Bombardment with deuterium edge plasmas, containing impurity W, produces a thick W layer on the C side of the limiter, whereas C implanted in the W side is strongly sputtered due to impact of most constituent D ions.

Tungsten coatings of 0.5 and 1 mm thickness were successfully deposited by the vacuum plasma spraying (VPS) technique on carbon/carbon fiber composite (CFC), CX-2002U and isotropic fine grained graphite, IG-430U. High heat flux experiments by irradiation of electron beam with uniform profile were performed on the coated samples in order to prove the suitability and load limit of such coating materials. The cross-sectional composition and structure of the interface of VPS–W and carbon material samples were investigated. Compositional analyses showed that the Re/W multi-layer acts as diffusion barrier for carbon and suppresses tungsten carbide formation in the VPS–W layer at high temperature about 1300°C. Microstructure of the joint interface of the sample changed in the case of a peak temperature of about 2800°C. The multi-layer structure completely disappeared and compositional distribution was almost uniform in the interface of the sample after melting and resolidification. The diffusion barrier for carbon is not expected to act in this stage.

The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma–material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R&D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R&D work is urgently needed.

The Large Helical Device (LHD) experiments have started after a construction period of eight years, and two experimental campaigns were performed in 1998. The magnetic field was raised up to 2.75 T at a magnetic axis position of 3.6 m at the end of the second campaign. In the third campaign, started in July in 1999, the plasma production with ECH of 0.9 MW and auxiliary heating with NBI of 3.5 MW have achieved an electron temperature of 3.5 keV and an ion temperature of 2.4 keV. The maximum stored energy has reached 0.75 MJ with an averaged electron density of 7.7×1019 m-3 by hydrogen pellet injection. The ICRF heating has sustained the plasma for longer than 2 s and the initial stored energy of the NBI target plasma has increased from 0.27 MJ to 0.335 MJ. The major characteristic of the LHD plasma is the formation of the temperature pedestal, which leads to some enhancement of energy confinement over the ISS95 scaling law. The confinement characteristic is gyro-Bohm and the maximum energy confinement has reached 0.28 s. The LHD has also shown its high potentiality for steady-state operation by realizing a 22 s discharge in the second campaign.

CX-2002U and IG-43OU coated by VPS-W were developed to be as a light high-Z plasma-facing material. After brazing them on OFHC blocks using a Ti foil, their thermal response and thermal fatigue properties were examined with active cooling. No cracks and no exfoliation occurred on the W surface and the braze interface even after 160 cycles of heat load for 20 s at 10 MW/m2. This result indicates that the Ti-brazing is a possible alternative to Ag-brazing for joining carbon to Cu. Heat load resistance of the VPS-W coated CX-2002U/OFHC was much better than the VPS-W coated IG-430U/OFHC due to the excellent thermal conductivity of CX-2002U. VPS-W coated CFC/OFHC is a potential candidate for a high heat resistance armor material on plasma facing components.

Atom-order void formation in a defective crystalline material is studied by using molecular dynamics method (MD). Infinite long cylinder, which is constructed with nickel atoms with a line of vacancies, is subjected to multiaxial tensile strain field by moving periodic boundary and the atoms on the outer surface of the cylinder. When the load exceeds a critical value, sudden appearance of the void is observed and it develops rapidly. The developed void does not disappear by only unloading and relaxation, in spite of the system with the void has higher potential energy than that without void. The biaxial or the triaxial load is necessary to the atom order void formation. Moreover, the results by the MD simulations are compared with theoretical solution for nonlinear elastic solid.

The Large Helical Device is the world's largest heliotron type helical system, with the plasma confining magnetic field being generated by only external superconducting coils. One of the main objectives of the LHD project is to sustain high temperature plasmas for a long time in steady state. The plasma vacuum vessel and the divertor are water cooled, and a heat load of 3 MW can be removed continuously. The NBI, ECH and ICRF heating systems, diagnostic instruments and data acquisition system are designed for long pulse operation. The present status of these systems and the recent experimental results of long pulse operation are reviewed. A steady state discharge with NBI was obtained for 35 s. The ECH discharge duration was extended to 120 s with a duty factor of 95%. Plasma sustainment by ICRF alone was achieved for 2 s. The performance of these long pulse operations is summarized.

The first experiment of the ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device (LHD) was carried out at the end of 1998. The LHD is a large superconducting heliotron device and its first plasma was produced in March 1998. During the ICRF heating experiment, a maximum 300 kW/0.2 s of ICRF power was injected into the LHD plasma by using a pair of loop antennae. This paper reports on the installation of the loop antennae, the results of antenna coupling and the first heating experiments. The antennae are designed to operate in the steady state and to change their distance from the plasma by 0-15 cm. In the experiment, the antenna resistance coupled with the plasma was measured by changing the distance between the last closed flux surface and the launcher front from 9 cm to 5 cm. The resistance was almost doubled by decreasing the distance. The target plasma was produced by the second harmonic electron cyclotron heating (ECH) of 84 GHz gyrotrons at a magnetic field of 1.5 T and a low plasma electron density of less than 1 × 1019 m-3 . Therefore, the low coupling resistance limited the maximum injected power to less than 300 kW. The heating efficiency and heating species were varied by the minority ion gas-puffing rate. The heating characteristics were compared with a one-dimensional full-wave analysis code, and the experimental results were consistent with wave damping analysis. For the optimum condition of the minority hydrogen gas-puff ratio, the plasma internal energy increased from 13 kJ to 26 kJ by almost the same power as the ECH power.

In March 1998, the LHD project finally completed its eight year construction schedule. LHD is a superconducting (SC) heliotron type device with R = 3.9 m, ap = 0.6 m and B = 3 T, which has simple and continuous large helical coils. The major mission of LHD is to demonstrate the high potential of currentless helical-toroidal plasmas, which are free from current disruption and have an intrinsic potential for steady state operation. After intensive physics design studies in the 1980s, the necessary programmes of SC engineering R&D was carried out, and as a result, LHD fabrication technologies were successfully developed. In this process, a significant database on fusion engineering has been established. Achievements have been made in various areas, such as the technologies of SC conductor development, SC coil fabrication, liquid He and supercritical He cryogenics, development of low temperature structural materials and welding, operation and control, and power supply systems and related SC coil protection schemes. They are integrated, and nowadays comprise a major part of the LHD relevant fusion technology area. These issues correspond to the technological database necessary for the next step of future reactor designs. In addition, this database could be increased with successful commissioning tests just after the completion of the LHD machine assembly phase, which consisted of a vacuum leak test, an LHe cooldown test and a coil current excitation test. These LHD relevant engineering developments are recapitulated and highlighted. To summarize the construction of LHD as an SC device, the critical design with NbTi SC material has been successfully accomplished by these R&D activities, which enable a new regime of fusion experiments to be entered.

Long-pulse neutral beam injection heating has been achieved in the large helical device (LHD). Two different confinement states are observed for different averaged densities in the long-pulse plasmas. A quasi-steady-state plasma was sustained for 21 s with an injection power of 0.6 MW, where the central plasma temperature was around 1 keV with a line-averaged electron density of 0.3 × 1019 m-3 . The discharge duration can be so extended as to keep the plasma properties in the short-pulse discharge. The energy confinement time is nearly the same as that of the short-pulse discharge, which is 1.3 times as long as the international stellarator scaling ISS95. At higher densities, a relaxation oscillation phenomenon, observed as if the plasma would breathe, lasted for 20 s with a period of 1-2 s. The phenomenon is characterized with profile expansion and contraction of the electron temperature. The density oscillation is out of phase with the temperature oscillation and is related to the density clamping phenomenon. The observed plasma properties are shown in detail for the `breathing' oscillation phenomenon. Possible mechanisms for the breathing oscillation are also discussed, with a view of the screening effect near the last closed magnetic surface and the power balance between the heating and the radiation powers. The long-pulse heating results indicate unique characteristics of the LHD where no special feedback stabilization is required due to absence of disruption and no need for current drive.

Saccharomyces cerevisiae alpha-mannosidase (Ams1) is a cargo protein that is transported to the vacuole by the cytoplasm-to-vacuole targeting (Cvt) pathway during conditions of growth and by autophagy during conditions of starvation. After transport to the vacuole, Ams1 functions as a resident hydrolase. Ams1 has been overexpressed in the methylotrophic yeast Pichia pastoris, purified and crystallized in two crystal forms. Form I belongs to space group P2(1), with unit-cell parameters a = 145.7, b = 127.7, c = 164.0 A, beta = 101.5 degrees . Form II belongs to space group I222 or I2(1)2(1)2(1), with unit-cell parameters a = 127.9, b = 163.7, c = 291.5 A. Diffraction data were collected from these crystals to a resolution of 3.3 A for form I and of 2.6 A for form II using synchrotron radiation.

We have achieved long-pulse plasma heating using a negative-ion-based neutral beam injector (NBI) in the large helical device (LHD), where the confinement magnetic field is generated by only external superconducting coils. In the initial long-pulse experiments at lower power than that in short-pulse experiments, 80 keV–1.1 MW NBI heating lasted for 10 s with a little increase in the plasma density at the pulse end. Almost steady-state plasma heating was achieved for 21 s with 66 keV–0.6 MW NB injection. Plasma relaxation oscillation phenomena at a period of 1–2 s were also observed for 20 s. Above 1 keV plasma was easily sustained with a long-pulse NBI heating in LHD, without the current drive nor the disruption in tokamaks. Negative ion source operation was stable and the cooling water temperature rise of beam accelerator grids was nearly saturated with a temperature rise below 10 °C. For a higher power injection, the pulse duration is determined by the beam blocking, where the reionization loss is expon...

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns were carried out in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 1019m-3, central temperatures ranging from 1.4 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scaling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

Particle release from a W test limiter exposed to deuterium edge plasmas, containing impurities C and O, in the tokamak TEXTOR-94 is studied both experimentally and theoretically by means of Monte-Carlo simulations. The simulation model combines the dynamic composition change in the surface layer with the transport of released particles in the plasma. The release of W impurities from the limiter is strongly suppressed by prompt redeposition, whereas low- Z impurities (C and O) and D are much less influenced. Experimentally observed radial distributions of WI and CII line intensities in front of the limiter can be explained by physical sputtering of W and deposited C, respectively, in addition to high-energy reflection of impurity C ions. On the other hand, the CII line emission observed in front of graphite as a limiter is attributed to many low-energy (∼0.1 eV) atoms probably due to chemical sputtering by impact of D ions, which is strongly suppressed by deposition of high- Z impurities, such as W. The observed Dγ line emission indicates an important contribution of high-energy reflection to the release of D atoms from the W limiter, in addition to a large contribution of low-energy re-emission.

The release of deuterium and impurity atoms from a W test limiter exposed to the edge plasma of TEXTOR is studied both experimentally and by means of Monte Carlo simulations. Experimentally, spatial intensity distributions of D γ , CII, OII and WI line spectra emission are observed around the limiter, whereas the particle release from the limiter is simulated combining the TRIDYN model with a transport model of released particles in the plasma. Good agreement is found in spatial distributions between experimentally observed WI line intensity and simulated ionization events of physically sputtered W atoms. The observed D γ line emission is attributed to the re-emission of very low-energy (∼ 0.1 eV) D atoms, the energy of which depends on the limiter temperature. Low-energy C and O atoms are also observed probably due to chemical sputtering or surface reaction of implanted (deposited) impurities producing hydrocarbons and volatile oxides. The observed CII and OII line emissions at a distance more than about 1 cm from the limiter surface are influenced by the physical sputtering of impurities deposited from the background plasma on the W limiter.

The effect of oxygen contamination on hydrogen absorption and desorption behavior from a boron coating film has been studied. Oxygen atoms were implanted by glow discharge in an O2/He gas mixture until near saturation, into the boron film deposited by PCVD. The depth profile measurement by AES showed that O atoms were retained up to the depth of 20 nm. Hydrogen discharges were carried out to investigate the H absorption behavior. The capability of H absorption decreased for 30–50% compared to the pure boron film without O contamination. After the discharge, the depth profile of the oxygen atoms was not changed, which means that a stable oxide layer had formed. The reduction of the H absorption capability occurs probably because the formation of the boron oxide prevents H atoms from trapping in the form of B–H bonding. Most of the retained H atoms can be released by a heating up to 500°C with the O contamination. The required temperature for H evacuation is slightly higher than that for pure boron film. In addition, a small peak was observed at around 200°C. From these results, the applicability of boronization to future long term discharges was discussed, in which the boron film saturates with O contamination.

Reflection of ions at a tungsten target was investigated experimentally in a linear plasma device. The tungsten target plate was set perpendicular to the magnetic field, and irradiated by hydrogen and helium plasma. The experiment measures a current as a function of sheath potential for a fixed angle. The reflected neutral particles from the target plate are measured by neutral-negative charge convertor made of Cu. The experimental results show good agreement with the results calculated with TRIM Monte-Carlo simulation code. A possibility to estimate the ion temperature is shown through the sheath potential dependence of the ion reflection in the low energy range.

Carbon sheet pump (CSP), which has been developed for reduction of hydrogen recycling in the large helical device (LHD) of National Institute for Fusion Science, is applied to an actual plasma device for the first time to evaluate the pumping characteristics. A small scale of pump module is manufactured and installed in the GAMMA-10 tandem mirror device and the module is exposed to charge-exchange neutrals emitted from ICRF-heated hot-ion plasmas. Temporal evolution of the pressure is measured with a fast ionization gauge installed in the test chamber during the plasma discharge and a significant difference between CSP-on and CSP-off is observed under good reproducibility. The dependence on the ICRF power is investigated and the pressure difference between on and off is found to increase with the ICRF power. These phenomena indicating the pumping ability of CSP are observed to continue during more than 300 plasma shots, which shows the effectiveness of CSP in controlling hydrogen recycling. The result of thermal desorption experiments of CSP after being exposed to charge-exchange neutrals is also described.

A new diagnostic method to measure ion energy distribution using ion reflection on the solid surface, one of the processes of plasma-wall interactions, is proposed. It is well known that the energy distribution of reflected particles is related to the energy distribution of incident ions. This method is used in relatively Iow ion temperature region such as edge plasma in fusion experimental devices. Incident ions to target plate which is inserted in the edge plasma are mainly reflected as neutral particles, and time-of-flight method is utilized to measure energy distribution of reflected particles. Energy distribution of incident ions is deduced by fitting with calculation using Monte Carlo simulation code TRIM.SP. The sheath potential in front of the target can also be obtained with this method.

Tungsten coatings of 0.5 and 1 mm thickness were successfully deposited by the vacuum plasma spraying technique on carbon/carbon fiber composite (CFC), CX-2002U, and isotropic fine grained graphite, IG-430U. High heat flux experiments by irradiation of electron beam with uniform profile were performed on the coated samples in order to prove the suitability and load limit of such coating materials. Heat load properties, gases emission, surface modification and structure changes of cross-section of the samples were investigated. Cracks on the surface and exfoliation between the joint interface of the samples were not formed below the melting point. These results indicated that the thermal and adhesion properties between the substrate and coatings were good under high heat flux. Microstructure of the joint interface of the sample was changed in the case of a peak temperature at about 2800°C. Many cracks and traces of melted tungsten flow were observed on the surface after melting and solidification. Large cavities were also formed inside the resolidified tungsten layer.

A thin boron film is attractive as a deuterium/tritium free wall, and as a protecting layer against impact of energetic charge-exchange neutrals in future fusion devices with long pulse operation. New experimental evidence is given for desorption of hydrogen isotopes from these films at relatively low temperature. Most hydrogen atoms in a boron-coated layer are re-emitted to the plasma side below 400°C without penetration into the substrate of stainless steel. The maintainability of a thin boron layer during a long pulse operation may be a problem. Boron atoms are hardly removed by pumping because their hydrides are easily disintegrated and redeposited. Gross migration of boron atoms inside the vessel is a concern. A condition required for avoiding the migration is discussed.

The LHD experiment has just begun. A scenario is presented for LHD divertor experiments. It includes development of LHD divertor components particularly efficient pumping system, local island divertor as a closed pumped divertor, simultaneous achievement of H-mode and radiative cooling (SHC operation) as an H-mode approach in the helical device, high temperature divertor plasma operation for enhancement of the energy confinement.

The aim of these investigations is the description of the power flux on roof-like limiters near the LCFS. Here, the specific subject is the toroidal pump limiter ALT-II of TEXTOR-94. Similar to other plasma facing objects, the surface of ALT-II is shaped to enhance the plasma wetted area; this is achieved by reducing the angle of incidence of the magnetic field to less than 1? for the first 10?mm of the SOL. This small angle of incidence enhances all effects of toroidal non-uniformity as given, for example, by magnetic field ripple. Extensive modelling explains well the observed heating pattern on the limiter surface due to the ripple effect. In contrast to the expectations from density and temperature distributions in the SOL and at the edge of the confined region, an excessive power density is deposited on the first few millimetres near the tip of the limiter roof. Physical effects which could cause this phenomenon are discussed.

Atg16 is a scaffold protein that interacts with Atg12-Atg5 protein conjugates via its N-terminal domain and self-assembles via its coiled-coil domain, thus forming a multimeric Atg12-Atg5-Atg16 complex that is essential for autophagy. The coiled-coil domain of Atg16 was expressed, purified and crystallized. The crystal belonged to space group P4(1)2(1)2 or P4(3)2(1)2, with unit-cell parameters a = 127.7, c = 77.8 A. Self-rotation functions and volume-to-weight ratio values suggested that the crystal may contain six molecules per asymmetric unit. Since the domain does not contain a methionine residue, selenomethionine-labelled crystals were prepared with a leucine-to-methionine substitution in the coiled-coil domain and these crystals were used for the collection of single-wavelength anomalous dispersion data to 2.5 A resolution.

The Atg12–Atg5 conjugate was prepared by in vivo reconstitution and was crystallized with Atg16 using the free-interface diffusion method., Autophagy mediates the bulk degradation of cytoplasmic components in lysosomes/vacuoles. Five autophagy-related (Atg) proteins are involved in a ubiquitin-like protein conjugation system. Atg12 is conjugated to its sole target, Atg5, by two enzymes, Atg7 and Atg10. The Atg12–Atg5 conjugates form a multimeric complex with Atg16. Formation of the Atg12–Atg5–Atg16 ternary complex is crucial for the functions of these proteins on autophagy. Here, the expression, purification and crystallization of the Atg12–Atg5 conjugate bound to the N-terminal region of Atg16 (Atg16N) are reported. The Atg12–Atg5 conjugates were formed by co-expressing Atg5, Atg7, Atg10 and Atg12 in Eschericia coli. The Atg12–Atg5–Atg16N ternary complex was formed by mixing purified Atg12–Atg5 conjugates and Atg16N, and was further purified by gel-filtration chromatography. Crystallization screening was performed by the free-interface diffusion method. Using obtained microcrystals as seeds, large crystals for diffraction data collection were obtained by the sitting-drop vapour-diffusion method. The crystal contained one ternary complex per asymmetric unit, and diffracted to 2.6 Å resolution.

The Tob/BTG family is a group of antiproliferative proteins that contain two highly homologous regions named Box A and Box B. These proteins all associate with CCR4-associated factor 1 (Caf1), which belongs to the ribonuclease D family of deadenylases. The antiproliferative region of human Tob (residues 1-138) and intact hCaf1 were co-expressed in Escherichia coli, purified and successfully cocrystallized. The crystal belongs to the tetragonal space group I422, with unit-cell parameters a = b = 150.9, c = 113.9 A, and is estimated to contain one heterodimer per asymmetric unit. The crystal diffracted to around 2.6 A resolution.

The Atg1/ULK complex functions as the most upstream factor among Atg proteins to initiate autophagy. ATG101 is a constitutive component of the Atg1/ULK complex in most eukaryotes except for budding yeast, and plays an essential role in autophagy; however, the structure and functions of ATG101 were largely unknown. Recently, we determined the crystal structure of fission yeast Atg101 in complex with the closed HORMA domain of Atg13, revealing that Atg101 is also a HORMA protein with an open conformation. These 2 HORMA proteins play essential roles in autophagy initiation through recruiting downstream factors to the autophagosome formation site.

Tungsten coatings of 0.5 and 1.0 mm thickness were successfully deposited by the vacuum plasma spraying technique (VPS) on carbon/carbon fiber composite, CX-2002U, and isotropic fine grained graphite, IG-430U. High heat flux experiments were performed on the coated and non-coated samples in order to prove the suitability and load limit of such coating materials. The electron beam irradiation experiments showed that there was little difference in temperature increases among CX-2002U and the coated materials below surface temperature of 2200°C. These results indicated that thermal and adhesion properties between the substrate and coatings were good under high heat flux. A few cracks with a width of 15 μm were formed from the surface to the bottom side of the all coated samples, but plastic deformation and microcracks due to grain growth by recrycrallzation were not observed below a surface temperature of about 2200°C. The cracks are expected to be formed due to local thermal stress produced by spot-like beams.

The main feature of FFHR is force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety. Collaboration works based on the LHD project have made great progress in the reactor studies by focusing on engineering aspects of the high magnetic field and Flibe system design. Encouraging positive results are shown on ignition access, mechanical stress in coils supporting structures, improvement in the blanket system including materials selection and tritium recovery. Critical issues on fundamental safety analysis and maintainability of reactor components are also discussed, and many subjects are pointed out as future works.

In the reactor design, many components are installed in a narrow space between the superconducting coil and the plasma. In the large helical device (LHD) configuration, the plasma axis is shifted inwards relative to the centre of the two pairing helical coils, and so the space available for thermal insulation, the vacuum vessel, cooling pipes and the first wall is very narrow. The LHD plasma vacuum vessel maintains good vacuum conditions, removes heat from the plasma, and protects superconducting coils from the plasma heat. The vessel is designed to have enough rigidness for stress deformation. In the construction, to facilitate assembly with the helical coils which were completed earlier, the vacuum vessel was fabricated in sections of many small segments. Design and construction of the LHD plasma vacuum vessel is described. These technologies are also quite important to reactor design.

To develop the divertor plates for the large helical device (LHD), several kinds of brazed modules (BM) and mechanical jointed modules (MJM) have been fabricated and their thermal properties evaluated using an electron beam facility. The MJM and BM will be used as divertor plates in low heat flux (≈5 MW m−2) and high heat flux (≈10 MW m−2) phases, respectively. Results of the high heat flux tests indicate that some types of BMs satisfy the requirements for the divertor plates. For MJM, the efforts to improve the thermal properties are now being made so as to satisfy the divertor plate requirements. To optimize the design and assembly procedures, divertor mockups of the MJM have been fabricated and installed inside a vacuum vessel mock-up of LHD. The developments of these divertor components and the results of the high heat flux tests are reviewed.

Dynamic behavior of hydrogen atoms in boron films, which is one of candidate materials of the first wall of fusion devices, has been studied. Hydrogen absorption behavior was investigated with a glow discharge in hydrogen. After a strong absorption in the beginning of the discharge, hydrogen atoms were slowly and continuously absorbed for 10 h without saturation. The slow absorption is due to deeper migration of hydrogen atoms into the film, enhanced by the ion bombardment. A helium glow discharge causes both ion induced desorption and the deeper migration, and thus, hydrogen atoms are accumulated in the film when the H 2 and He discharges are repeated alternately. Most of the retained hydrogen atoms were released by heating up to 400°C. A calculation result based on a simple model was compared with the experimental results. A recombination coefficient was obtained from the analysis of the experimental transient release of the H atoms just after the H 2 discharge. The time behavior at the initial phase of the He discharge was reproduced fairly well with use of a smaller cross-section for the He + ion induced detrapping than that for H 2 + .

An enhancement of antiprotons produced in p+d reaction in comparison with ones in p+p elementary reaction is investigated. In the neighborhood of subthreshold energy the enhancement is caused by the difference of available energies for antiproton production. The cross section in p+d reaction, on the other hand, becomes just twice of the one in elementary p+p reaction at the incident energy far from the threshold energy when non-nucleonic components in deuteron target are not considered., LaTeX,7 pages with 5 eps figures

Five new ether-soluble resin glycosides were isolated from whole plants of Ipomoea stolonifera. Their structures have been determined on the basis of chemical and spectral data. Similar to the resin glycosides previously isolated, all of them are monomers of a jalapinolic acid tetra- or penta-glycoside in which the sugar moiety is partially acylated by organic acids and also combined with the carboxy group of the aglycone to form a macrocyclic ester structure.

Background Stage migration in gastric cancer confounds establishment of standard treatment according to stage. Methods To determine how closely lymph nodes should be examined to report correct staging, 402 node-positive patients were evaluated retrospectively. A total of 23 233 lymph nodes were reviewed histologically and their maximum dimension was measured. Another 254 nodes from 12 patients were used to evaluate shrinkage after fixation and preparation of the histological slide. Results Metastasis was detected in 3142 nodes, 1163 with well differentiated tumours (WDTs) and 1979 with poorly differentiated tumours (PDTs). Mean(s.d.) size of metastatic nodes was 7·80(5·08) mm in all, 8·44(5·74) mm in WDTs and 7·42(4·62) mm in PDTs. Both positive and negative nodes shrank between 10 and 20 per cent during histological processing. If all nodes 5 mm or less in size when fixed are ignored 37·8 per cent of all metastatic nodes will be missed. Downstaging will occur in 14·9 per cent and 4·2 per cent of the cases if all nodes less than 6 and 4 mm respectively are ignored. Conclusion To keep the rate of stage migration caused by this factor below 5 per cent, all lymph nodes 4 mm or more in size (5 mm when fresh) should be retrieved and examined.

Boron film coated on plasma facing walls has been utilized to reduce the oxygen impurity level by the gettering action. The boron film is also useful to reduce the hydrogen recycling. In this study, the boronization was conducted by a dc glow discharge with a mixture gas of diborane and helium both for a graphite and a stainless steel (SS) liners. After the boronization, the oxygen glow discharge was carried out to evaluate the gettered oxygen amount. The state of the oxygen in the surface was also examined. The gettered oxygen amount in the case of the graphite liner was about twice larger than that in the case of the SS liner. The oxygen was trapped in the depth range from the top surface to 100 nm or from the top surface to 20–30 nm in the case of graphite or SS, respectively. The oxygen was observed to be chemically bonded with the boron. After the oxygen discharge, the helium discharge was conducted to recover the oxygen gettering ability. After the helium discharge, the oxygen discharge was again carried out. The gettered oxygen amount in the case of graphite was comparable with that in the case of SS.

The blanket design for a force-free helical reactor (FFHR) is presented, which is a demo-relevant heliotron-type D-T fusion reactor based on the first all-superconducting-coils device, LHD (large helical device) under construction in NIFS at present. For the goal of a self-ignited reactor of 3 GW thermal output, the design parameters at the first stage for concept definition of FFHR have been investigated. The main feature of FFHR is a force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use a high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt FLiBe as a self-cooling tritium breeder for mainly safety reasons owing to the low tritium inventory, low reactivity with air and water, low pressure operation, and low MHD resistance compatible with a high magnetic field. In particular, as common issues in fusion reactors, the FLiBe blanket system in FFHR is expressed in detail by showing engineering possibilities to overcome key issues on tritium permeation, material corrosion, heat transfer, operation pressure, etc. The basic design for maintenance and repair of the blanket is also discussed.

Studies directed towards the reduction of particle and heat fluxes to plasma facing components by the application of ponderomotive forces generated by radio frequency (RF) are being conducted in TEXTOR. A modified poloidal limiter is used as an antenna with up to 3 kW of RF power; the data obtained show that the plasma is repelled by the RF ponderomotive potential. The density is reduced by a factor of 2–4 and the radial decay length is substantially altered. The density near the limiter decays exponentially with RF power. The electron temperature profile changes, with the decay length becoming longer (almost flat) during the RF. The temperature in the scrape off layer (SOL) increases and its increase is roughly proportional to the RF power until it saturates, suggesting that the heating efficiency drops with power, and that improved performance is to be expected at higher powers.