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This study developed a new correction method to reduce errors in temperature measurement caused by temperature-dependent emissivity for ITER divertor infrared (IR) thermography. In order to develop the method, we measured the spectral emissivity of tungsten from 200 °C to 1900 °C in the wavelength range of 1.5 μm to 4.5 μm and evaluated temperature measurement error of the two-color method. The results indicate that temperature-dependent emissivity is a major factor in temperature measurement errors, which suggests that accurately measuring temperature by using the conventional two-color method would prove difficult. The new correction method uses two emissivity inputs at different temperature points, in contrast to the conventional method, which uses emissivity measured at a single temperature point. Using this new correction method with a suitable wavelength combination for the two-color method, temperature measurement error was reduced to less than 5% at 1900 °C.

An important issue for ITER divertor infrared (IR) thermography (IRTh) is that changes in the emissivity of tungsten divertor targets resulting from depositions; erosions; and dependences on temper...

This study developed a new heat flux reconstruction code based on the dual reciprocity boundary element method for the International Thermonuclear Experimental Reactor (ITER) divertor infrared (IR) thermography system. To use divertor heat flux reconstruction in ITER, we modeled the boundary condition between the coolant pipe inner wall and the coolant based on the temperature-dependent heat transfer coefficient and also considered the temperature dependence of tungsten thermal properties. Using this reconstruction code, we evaluated the sensitivity of the input data errors, divertor coolant temperature, and surface temperature errors on the accuracy of heat flux reconstruction by using simplified heat flux profiles, which are spatially uniform on the top surface of the monoblocks. This heat flux reconstruction method demonstrated that highly accurate heat flux reconstruction can be achieved for high heat flux cases in ITER; however, further studies are needed for low heat flux reconstruction.

The harsh environment in the ITER device presents many diagnostic challenges. A proposed design engineered at the Japan Atomic Energy Agency (JAEA) is of an optical alignment system installed on mirrors behind double vacuum windows. Due to the harsh environment, a piezo-motor is used as an actuator and a sliding bearing is used as a pivot. The required torque of the actuator is estimated. Trial results indicate that the actuator as described fulfills requirements. A prototype of a key optical component, the micro-retro-reflector array, is produced through the electro-forming method. Measured reflectivity yields a sufficient signal for in situ calibration. A Cassegrain telescope prototype is designed and produced using simple spherical mirrors and achromatic lenses made of silica and CaF 2 . The measured focal length concurs with design objectives. Irradiation tests are necessary to verify the capabilities of these optical devices.

The technological challenges associated with preparing and installing diagnostic systems on ITER are summarized and the solutions that have been developed to overcome them are described. Areas where further developments are needed are highlighted.

The concept of multiplexing thermography is applied to the design of the divertor thermography system for International Thermonuclear Experimental Reactor (ITER). The combination of the front mirror with multiellipticity and a Czerney–Turner spectrometer with a 0.2 mm pitched multichannel detector enables a spatial resolution of 3 mm and a time resolution of 20 μs above a target temperature of 300 °C to be achieved. This should be sufficient to measure ELM heat fluxes to the targets in ITER. To satisfy the measurement requirement, it is very important to keep an accurate alignment around the optical axis against movement of the vessel during the plasma discharges. Several key engineering problems, such as the survivability of components against mirror coating by redeposited divertor material, remain to be solved. Potential solutions have been identified.

ITER diagnostic equipment is integrated in six equatorial and 12 upper ports, five lower ports, and the 16 divertor cassettes directly in front of them, and at many locations in the vacuum vessel. The integration has to satisfy multiple requirements and constraints and at the same time must deliver the required diagnostic performance. Design work has been carried out in problematical and interfacing areas for most diagnostics. The main diagnostic support components, the port plugs and divertor port racks, comprise common structural elements with dedicated modules from several diagnostics. The scope of the engineering task is summarized and an overview of the integration of the diagnostic equipment is given. The engineering work to date represents the input to the port engineering tasks of the construction phase with common or repeated elements being assessed for feasibility, resolving issues, and identifying show stoppers. It has been dominated by allocation and spatial integration in the tokamak. The work that follows will be more oriented towards the acceptability for operation, suitability for function, and conformity for licensing.

The main plasma regions of ITER ‐ the core, the edge, the scrape-off layer, and the divertor ‐ will be probed by an extensive array of spectroscopic diagnostics covering the visible to X-ray wavelength range. Plasma parameters will be determined including impurity species/density/input-flux, ion temperature, helium density, fueling ratio, plasma rotation, effective ionic charge and safety factor q. The measurements will be used for plasma control and in studies to understand and improve the performance of ITER. Both passive and active techniques will be employed. A diagnostic neutral beam (DNB) (~100 keV) will be installed for Charge Exchange Recombination Spectroscopy. Motional Stark Effect measurements (for q profile) can in principle be made using both the heating neutral beam (1 MeV) and the DNB and both possibilities are under consideration. Diagnostic components, such as mirrors, windows, and optical fibers etc, mounted close to the plasma will experience higher levels of radiation due to neutron, gamma ray and particle irradiations than in present devices. Potentially their performance characteristics can be degraded and so the materials of the components have to be carefully selected and mitigating methods adopted where possible. This paper presents an overview of the ITER spectroscopic diagnostic systems and the details of some of the individual systems.

Heat and particle control has been studied under the reactor-relevant high-power heating in the large tokamak of JT-60U with an open divertor and progressively a W-shaped pumped divertor. Heat and particle control is crucial for reduction in heat load onto the divertor plates, control of density in the main plasma, effective exhaust of helium ash, and reduction in impurity contamination. For the reduction of heat load, radiative divertor concept was developed based on understanding of heat and particle transport in scrape-off layer and divertor plasmas, which contributed to establishment of divertor concept in ITER. With argon injection, the total radiation loss power reached up to 80% of the net heating power with high confinement of HH{sub y2} {approx} 1, where HH{sub y2} is a confinement enhancement factor over the IPB98(y,2) ELMy H-mode scaling, at high density of 80% of the Greenwald density in the ELMy H-mode plasma. For the density control, the dependence of particle confinement on plasma parameters was systematically studied with two confinement times for center- and edge-fueled particles, which enabled discussion of density controllability. Core fueling using a high-field-side pellet injection extended the operation range of high confinement (HH{sub y2} {approx} 1) from 60 to 70% ofmore » the Greenwald density in the high {beta}{sub p} ELMy H-mode plasma. Efficient helium ash exhaust of {tau}{sub He}*/{tau}{sub E} = 2.8 was demonstrated in the ELMy H-mode plasma with the pumping from the private flux region, which is the same pumping geometry as that in ITER design. Reduction in Z{sub eff} by puff-and-pump scheme was demonstrated, and chemical sputtering yields were estimated with the consideration of not only methane but also heavier hydrocarbons. Their sputtering yields showed strong dependence on the wall temperature and weak dependence on the particle flux. The measured profiles of C II and C IV line intensities were well reproduced by the Monte Carlo impurity transport simulation code (IMPMC code). The estimation of sputtering yields and development of the simulation code enabled reliable predictions for impurity behavior in a fusion reactor.« less

The chemical sputtering yields of CH4/CD4 and C2Hx/C2Dx have been measured at the divertor plates of JT-60U. Spectroscopic measurements for CH/CD and C2 spectral bands are applied to estimate the CH4/CD4 and the C2Hx/C2Dx flux. At the surface temperatures of 380, 440 and 560 K, the CH4 yield is, respectively, ~0.8%, 1-2% and 2-3%, the C2Hx yield 1-2%, 3-4% and 4-5%, and the total sputtering yield by hydrogen ions 3-4%, ~8% and ~10%. With increasing ion flux to the divertor plates (Γion), the sputtering yields (Y) decrease, i.e. Y∝Γion(-0.05 to -0.40). With increasing electron temperature (Te), the sputtering yields increase, i.e. Y∝Te0.5. It is concluded from the result of regression analysis of Y∝Te0.5 that the negative dependence of the yields on the ion flux is attributed to the incident ion energies to the carbon plates. The ratio of the sputtering yields by deuterium ions and hydrogen ions is estimated to be ≥1.5 based on the ion flux measurement by Hα/Dα intensity. The C2Hx/C2Dx sputtering yield accounts for ~80% of the total number of sputtered carbon atoms.

The divertor geometry of JT-60U was modified in 1998 to add a pumping slot near the outer strike point for improved pumping. In the both-side pumping divertor, the x-point MARFE onset density is ∼10% higher with pumping than without pumping. Radiated power near the x-point is decreased by pumping, and the divertor pumping is effective in suppression of the x-point MARFE. The effect of outer pumping is investigated in L-mode plasmas. The onset densities of the x-point MARFE are the same in both divertors. The fraction of total radiated power over absorbed power at the x-point MARFE onset is ∼20% higher in the both-side pumping divertor than in the inside pumping divertor, but it does not depend on the onset density and on a distance between the private dome and the outer leg. Carbon penetration from the divertor region to the main plasma has a weak dependence on the gap-out in the both-side pumping divertor.

Transport coefficients of impurities in reversed shear plasmas are evaluated and compared with those in ELMy H-mode plasmas. The diffusivity is significantly reduced (∼1/10) at the internal transport barrier. An inward pinch velocity of −3 m/s is also observed at the ITB. The inward pinch velocity increases as charge number of impurity increases. Effects of Ar seeding to the main plasma are investigated in ELMy H-mode. Argon seeding maintain improved confinement HITER89=1.4, the clear edge pedestal, the radiation fraction of ∼80% and the divertor detachment up to 70% of Greenwald limit. Moreover, argon seeding reduces type I ELM frequency less than half, but the heat load of each ELM on the outer target plate does not change. Therefore, the reduction of energy losses due to ELMs probably causes the improvement of edge confinement in Ar seeded cases, though the radiated power inside the separatrix increases.

A radiative divertor experiment with neon gas puff was carried out with the aim of investigating impurity behaviour, after the open divertor was modified to the pumped W-shaped divertor. To evaluate neon radiation, we analysed lines from the neon ion measured with a VUV spectrometer in the divertor plasma. As a result, lines from Ne IV-Ne VIII were identified. By combining the measurement of an absolutely calibrated multichannel interference filter spectrometer with that of the VUV spectrometer, the radiation loss through neon line emission was estimated in the divertor plasma. In the case of pumping-off, the detached plasma evolved into multifaceted asymmetric radiation from the edge (MARFE). It was observed that the line intensities of highly ionized neon (Ne VII, Ne VIII) increased simultaneously with the formation of the MARFE and furthermore increased after the MARFE formation. It is considered that after the divertor plasma was detached, the plasma flow velocity and the friction force were weaker in the case of pumping-off than in the case of pumping-on and more impurities moved toward the X-point region. These observations imply impurity back flow from the divertor upstream by the reduction of friction force.

Experimental results related to the divertor geometry such as divertor plasma detachment, neutral transport and plasma energy confinement, were compared in the open and W-shaped divertors. The ion flux near the outer strike point was larger than in the open divertor, and the electron temperature at the target, Tediv, was reduced. Divertor detachment and x-point MARFEs occurred at n e 10–20% lower than that for the open divertor. Although the leakage of neutrals from the divertor to the main chamber decreased, a neutral source in the main chamber due to an interaction of the outer scrape-off layer (SOL) plasma to the baffle plates became dominant above the baffle. Degradation in the enhancement factor of the energy confinement was observed similarly in the open and W-shaped divertors. The neutral density inside the separatrix was estimated to be a factor of 2–3 smaller, which did not affect the energy confinement.

The divertor geometry of JT-60U was modified from an open type to a w-shaped pumped configuration. This paper discusses behavior of carbon impurity in the w-shaped divertor. Carbon generation by chemical sputtering is important in deuterium plasmas. The reduction of carbon generation in helium discharges is consistent with this conclusion. CD-band intensities in the private flux region were smaller in the w-shaped divertor than in the open divertor, suggesting that the divertor dome is effective in decreasing chemically sputtered hydrocarbon molecules. The observed CD-band intensity was fairly reproduced by simulation using the IMPMC-code. When deuterium gas was puffed into the divertor, Zeff value in the main plasma became about 15% larger than with puffing from the main chamber. For reducing impurity in the main plasma, the gas puffing from the top of main plasma was effective.

Performance of inner leg pumping of JT-60U W-shaped divertor is presented. Inner leg pumping works well as the dome separates inner and outer divertor channels for neutral particles in the private flux region. In ELMy H-mode with NBI heating power of 18 MW, the operational regime where inner leg pumping is effective extends to the divertor radiation range of about 60% of SOL power. The fraction of inner particle flux pumped is estimated to be about 3%. It is qualitatively shown that a combination of main gas puff and inner leg pumping reduces carbon impurities in the main plasma, suggesting the existence of SOL flow effect on impurity reduction.

Emission of He I lines (668nm, 706 nm and 728nm) has been simultaneously observed in the JT-60U divertor plasma. From calculations using a collisional-radiative model, intensity ratios of these He I lines are considered to be useful to measure electron temperature and density. In order to evaluate the feasibility of the intensity ratio method for temperature and density measurements, the measured intensity ratios are compared with those calculated from the temperature and density measured with Langmuir probes. As a result, the measured intensity ratios are well reproduced by the calculations. The discrepancies in the ratios I (728nm)/I (706nm) and I (668 nm)/I (728nm) between the measurements and calculations are about 12%.

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200–1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high...

In order to investigate the behaviour of helium atoms in the divertor region of JT-60U, the spectral profile of a He I singlet line (667.82 nm) has been observed with a high-resolution visible spectrometer. In attached divertor plasmas, the spectral profile is narrow in front of the divertor plates, because the line is mainly emitted from low-energy helium atoms desorbed from the plates. In the upper stream of the divertor plasma, as the electron density increases, the profile becomes broader and the temperature of helium atoms corresponding to the broadening increases from 0.2 to 1.5 eV. In detached divertor plasmas, the temperature is 0.9-1.5 eV. The profiles in attached divertor plasmas have been reproduced by simulation with a neutral transport code considering elastic scattering by proton collision. The broadening of the profile is attributed to the elastic scattering. The elastic scattering affects helium concentration in the main plasma and helium pumping efficiency. In the simulation for an attached divertor plasma, the penetration probability of helium atoms from the divertor plates to the main plasma and the helium atom flux to the gap for pumping increase by 30% due to the elastic scattering.

The feasibility of installing a vacuum ultraviolet spectrometer at the divertor level on the International Thermonuclear Experimental Reactor is examined including the related engineering aspects. A narrow gap of 10 mm between divertor cassettes is used as an access for lines of sight that cover the area from the X point to the lower half of the divertor leg, and good diagnostic performance is predicted. Further work necessary to develop this configuration is identified.

In order to understand the recycling and emission processes of deuterium atoms, spectral profiles of the line emitted from the divertor region of JT-60U have been observed with a high-resolution spectrometer and analysed by simulation with a three-dimensional neutral particle transport code. The profile has been explained as composed of narrow and broad components; the narrow component is attributed to dissociative excitation and electron collisional excitation of the atoms produced by dissociation, and the broad component is attributed to electron collisional excitation of the atoms produced by reflection and charge exchange. In low-density plasmas, the simulated line profile agrees reasonably well with that observed, although the component attributed to the atoms reflected at the divertor tiles is overestimated by a factor of about two. Dissociative excitation from deuterium molecules and molecular ions plays an important role for the line intensity. The ratio of the line intensity to the deuterium atom flux for high-energy deuterium atoms, which are produced by the reflection and charge exchange, is reduced, because the fast atoms readily escape from the divertor plasma. The width of the narrow component in a low-density case corresponds to a temperature of deuterium atoms of 1.3 eV, and that in a high-density case corresponds to a temperature of 2.2 eV.

In JT-60U, a visible spectroscopic system with high wavelength resolution has been developed to observe Doppler broadening of spectral lines emitted from divertor plasmas. An echelle grating is used to obtain the high wavelength resolution and the nominal wavelength resolution is 5.3 pm. Using optical fibre arrays, the Littrow mounting and a charge-coupled device camera, it is possible to observe 10 sight lines simultaneously with a spatial resolution of 1.2 cm.

CD/CH-band intensities emitted from hydrocarbon molecules have been measured in the divertor region of JT-60U and the chemical sputtering yield of methane was estimated as a function of the surface temperature and the deuterium ion flux. The chemical sputtering yield increases with the surface temperature and decreases with increasing ion flux density in the L-mode plasmas. The B4C converted CFC tiles are installed in JT-60U and it is found that the chemical sputtering of B4C converted CFC tiles is suppressed in comparison to normal CFC tiles.

Spatial distributions of absolute line intensities of D I and C II-IV in the divertor region have been measured simultaneously with a 38-channel optical fibre array in JT-60U. The radiative losses in the divertor region have been derived from the intensities. In L-mode discharges without the MARFE, the fractions of the radiative losses due to deuterium and carbon are 36% and 64%, respectively. The carbon radiation is related to the carbon influx by a simple impurity model based on the confinement time of the carbon ion on the assumption of the physical sputtering. The power radiated due to carbon accounts for a large increase in the radiative losses by the MARFE.

A two-dimensional impurity code based on Monte Carlo techniques (IMPMC) has been developed, in order to study the impurity behavior in the divertor plasma. The model includes (1) impurity generation at the divertor plate, (2) ionization of sputtered neutrals, (3) parallel motion of impurity ions along field lines, (4) Coulomb scattering, (5) cross-field diffusion, and (6) atomic processes. This model is applied to the carbon impurity behavior in a JT-60U NB heated plasma. The impurity generation mechanism, the effect of the thermal force on impurity sheilding, and the contribution of charge exchange recombination to C 3+ ion density near the plates are clarified. The measured spatial profile of the CIV line indicates that the diffusion coefficient, D ⊥ is around 1 m 2 /s. The ion temperature measured by Doppler broadening of the CIV line is found to be shifted to lower temperature side due to the overspread distribution of C 3+ ions in the SOL by the thermal force.

B 4 C conversion-coated carbon fiber composites (B 4 C-converted CFC) have been installed on the outboard strike point of the divertor plate in JT-60U to form a complete toroidal ring. The plasma-surface interactions of the B 4 C-converted CFC have been investigated in high power divertor operation with neutral beam (NB) heating power of - 30 MW for 2 s. It is found that the carbon impurity is reduced without a significant increase of the boron impurity using a 100 μm thick B 4 C layer with a B /C ratio of 2.7. The oxygen impurity is reduced significantly by the evaporation of a 300 μm thick B 4 C layer with a B /C ratio of 3.8. A reduction in the detrimental effects on plasma-surface interactions with carbon-based materials has been demonstrated for more than 1000 shots by the use of surface-boronized carbon tiles in JT-60U

In JT-60U boronization using decaborane was carried out. Boronization reduced the oxygen impurity in OH discharges and shortened the wall conditioning after the vacuum vessel vent and consequently enabled JT-60U to produce clean plasmas easily except for NB heated plasmas. After boronization, particle recycling was reduced drastically in OH and NB discharges. High confinement plasmas were obtained including high β p mode and H-mode discharges. In the latest boronization part of divertor plates were replaced with B 4 C coated tiles with a B 4 C thickness ∼ 300 μm. After introducing B 4 C divertor tiles, an explosive generation of boron particles from the tiles was observed. By the combined effects of boronization with decaborane and boron generation from B 4 C tiles, oxygen impurity was so low that oxygen line signals were reduced to noise levels after the latest boronization. On the other hand, boron burst from the B 4 C tiles restricted the operation of JT-60U.

Helium (He) transport and behavior have been investigated during ELMy (edge-localized mode) H-mode using a helium beam in JT-60U. Helium transport coefficients in the ELMy H-mode plasmas ( H = 1.5) are found to be D A = 1–2 m 2 /s and C V = 1.0–1.5 when a neutral beam of 60 keV helium atoms is used for a source of central fuelling, and D A = 1 m 2 /s and C V = 1.0–1.5 when a short pulse He gas puff is used as peripheral fuelling. No helium accumulation is observed during ELMy H-mode. The behavior of helium in the divertor region has been studied with divertor spectroscopic measurement and compared with that in the main plasma. The He recycling flux in the divertor during ELMy H-mode is about one order of magnitude less than during L-mode with the same level of He gas fuelling.

In JT-60U, spectra of deuterium and impurities in divertor plasmas were observed with absolutely calibrated VUV spectrometers. Charge state distributions for C II-IV and O III-VI were obtained from the spectral line intensities. The Li-like ion was dominant for carbon and the B- and Be-like ions were dominant for oxygen. Radiative power losses from divertor plasmas were calculated from the spectroscopic measurements. For a neutral beam heated discharge with a heating power of 12 MW, the fractions of the radiative power losses due to deuterium, carbon and oxygen were 58, 39 and 3.4%, respectively

The impurity behavior and the remote radiative cooling have been studied in the JT-60U divertor discharges with high power NB heating. NB at a power level of 20 MW was injected for 2 s, and no carbon bloom has been observed. For the first time, the measurement and the calculation for the carbon influx from the divertor plates are compared quantitatively. The carbon generation mechanism can be explained by the sputtering with deuterium (physical sputtering), oxygen and carbon. It is found that the contribution of the chemical sputtering with deuterium is small and that of the sputtering with oxygen is important. Remote radiative cooling can reduce the heat flux onto divertor plates. The dependence of the radiative cooling power on the electron density and the safety factor is presented. The favorable operation regime for discharges with high power auxiliary heating is discussed based on the experimental results.

Improvement of both energy and particle confinement was observed in beam heated hydrogen plasmas in JT-60 when the heating power and the line averaged density exceeded thresholds. The radiation power from the divertor plasma increased to as much as 50% of the heating power. This regime was termed improved divertor confinement (IDC). The discharges with improvement of the energy confinement by up to 20% and enhanced divertor radiation lasted for several seconds in a quasi-steady state without significant impurity accumulation in the core. The threshold line averaged electron density for the IDC regime increased linearly with power and dropped with increasing safety factor. The improvement was observed only when the ion Del B drift was towards the X-point. The behaviour of light impurities and in-out asymmetries in particle recycling changed with reversal of the toroidal magnetic field. These findings are consistent with the theoretical model in which collisional transport near the separatrix is connected with the onset of the IDC regime

The first experiment simulating helium ash exhaust in a DT fusion reactor has been performed in the JT-60 tokamak with {ital L}-mode divertor discharges. He was injected with a 30-keV neutral beam to realize core fueling. Divertor neutral pressures of hydrogen and He increased in proportion to the cube of the electron density of the main plasma. The He enrichment factor was 0.5 at {ital {bar n}}{sub {ital e}}=6{times}10{sup 19} m{sup {minus}3} and increased in proportion to {ital n}{sub {ital e}}. In an {ital L}-mode plasma of a fusion reactor, thermalized {alpha} particles (He ash) generated in the core plasma can be easily exhausted with a pumping speed of several tens m{sup 3}/s.

The detailed design of the Impurity Influx Monitor (divertor) has been carried out to provide the measurement capability in the harsh environment such as higher irradiation levels of neutron, gamma‐ray and particles than in present devices. The in‐situ calibration system using a micro retro‐reflector array has been developed to monitor the sensitivity change of the optical system due to the environmental effects. The optical alignment system for the Monitor has been developed by using a dedicated optics for alignment in the collection optics for measurement.

Helium ash exhaust characteristics have been studied in the JT-60 lower divertor with ohmic heating (OH) and neutral beam heating (NBH) of PNB ≦ 18 MW and Ip = 1 MA. Helium fueling was done by gas puff. The divertor neutral pressures of hydrogen (H2) and helium were measured by a residual gas mass analyzer. The helium enrichment in the divertor was also measured. In OH discharges with helium gas puff, no buildup of helium neutral pressure was observed and helium was depleted. On the other hand, in NBH discharges with helium gas puff, although enrichment of helium is not observed, the neutral pressures of H2 and helium increase in proportion to . The density dependence of the helium neutral pressures in NBH discharges is promising for efficient helium ash exhaust in fusion reactors.

In JT-60, H-mode experiments with outer and lower divertors have been performed. In the outer divertor discharge, an H-mode similar to the modes observed in the lower/upper divertor discharges is obtained. Its threshold absorbed power and electron density are 16 MW and 1.8 × 1019 m−3. In the two schemes of combined heating with NB + ICRF and NB + LHRF, H-mode discharges are also obtained. Moreover, in the new configuration with the JT-60 lower divertor, H-mode phases with and without edge localized modes are obtained. The improvement in the energy confinement time in both divertor configurations is limited to values within 10%. The paper mainly presents the H-mode results of the outer divertor discharges. Also, typical results of the lower divertor discharges are shown for a comparison of the H-mode characteristics of the two configurations.

The function of measurements of VUV radiation emanating from the plasma in the divertor region are briefly summarized. The engineering constraints that would have to be satisfied in the implementation of VUV measuring systems for this region are identified. Two proposed configurations for the installation of a VUV spectrometer are considered and the further work that is necessary to develop these design options is identified.

The design of the divertor impurity monitoring system for ITER is now well advanced. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The expected impurities are carbon, tungsten, beryllium and copper originating from the divertor target plate and from the surface of the first wall in the main chamber. Neon and other impurity gases injected into the plasma for radiation cooling in the divertor will also be observed. The wavelength range is 200 nm to 1000 nm. This system, which is one of the most important diagnostic systems for plasma control, is included in the start-up set of ITER diagnostics1.

Infrared (IR) thermography is a useful diagnostic tool for measuring surface temperatures on divertor plates by using light emitted in the IR range. Japan plans to procure and develop divertor IR thermography for the ITER project. The system design, optical design, and evaluation of detected photons have already been conducted. All optical devices will be installed in the occupied space without interfering with other devices. The required spatial resolution of 3mm on divertor plates is satisfied by the designed optics. The gamma-ray dose rate at 106 s after shutdown of operations (shutdown dose rate) in the interspace is low (15 20 μSv/h). Two color detectors and a spectroscopic system are designed using mirrors with high reflectivity in the range 1 5 μm. From the evaluation of detected photons, sufficient photons at 3 and 5 μm can be detected at low emissivity (e = 0.2).

The effects of edge-localized mode (ELM) mitigation coils (ELM coils) on the loss of NBI-produced fast ions and fusion-produced alpha particles are investigated using an orbit following Monte Carlo code. The ELM mitigation coil field (EMC field) may cause a significant loss of fast ions produced by NBI on the order of 16.0–17.0% for a 9 MA steady-state ITER scenario. A significant transit-particle loss occurs in the case of the toroidal mode number n = 4 in which magnetic surfaces are ergodic near the plasma periphery. When the number of ELM coils in each toroidal row is nine, the main toroidal mode n = 4 is accompanied by a complementary mode n c = 5. Concerning the resonance of fast-ion trajectories, the anti-resonant surfaces of n = 4 are very close to the resonant surfaces of n c = 5 and vice versa. Since the effect of resonance on fast-ion trajectories dominates that of anti-resonance, a synergy effect of the main and complementary modes effectively enlarges the resonant regions. In a single n-mode EMC field, the resonant and anti-resonant regions are well separated. The peak heat load due to the loss of NB-produced fast ions near the upper ELM coils is as high as 1.0–1.5 MW m−2, which exceeds the allowable level in ITER. Rotation of the EMC field is essential for ITER to alleviate the local peak heat load. Most loss particles hit the inner side of the torus of the dome in the ITER divertor. The loss of alpha particles is also increased by the effect of the EMC field. The loss is still acceptably low at less than 1.0%.

The in situ calibration method for the impurity influx monitor (divertor) is experimentally examined. The total reflectance of the optical path from the focal point of the Cassegrain telescope to the first mirror is derived using a micro retroreflector array. An optical fiber with angled physical contact (APC) connectors reduces the return edge reflection. APC fibers and a multimode coupler increase the signal-to-noise ratio by about one order compared to that of triple-branched fibers and enable measurement of the wavelength dependence of the total reflectance of the optical system even after potential deterioration of mirror surfaces reduces reflectance.

The intensity of emission lines in the ITER divertor plasma is synthesized using a model of the edge and scrape-off layer plasma. The intensity of hydrogen isotopes and impurities observed with the divertor central optical system of the Impurity Influx Monitor (Divertor) is estimated. In both inner and outer divertor plasmas, ionizing and recombining plasma components are observed for C I–III, and the recombining plasma component is dominant for C IV–V. Line-of-sight integrated spectra are synthesized in the wavelength range 200–1000 nm using photon emission coefficients. In order to evaluate the carbon-impurity influx, identifying the plasma component, i.e. an ionizing or recombining plasma component, is important.

We are developing a spectroscopic diagnostics system in ultraviolet and visible wavelength regions for monitoring ITER divertor plasmas. An equivalent-size prototype of the optical components for viewing upper port fanarray chords has been assembled as a system to measure spectral throughput, i.e., etendu. Collisional-radiative models for HeI and CIV are used to estimate the emission line intensities of helium ash and carbon impurity ions in a divertor region of a burning plasma. The estimated line intensity of CIV, λ772.6 nm, satisfies ITER requirements for the time resolution of measurement of Ti. A numerical simulation of the computerized tomographic technique for various pairs of viewing fan arrays has been applied to the divertor plasma region to reconstruct a two-dimensional distribution. The optimized pair of viewing fan arrays resolves a model distribution with a reasonable spatial resolution. We measure the reflectance of surfaces of carbon-fiber-composite and tungsten blocks, which make up the plasma-facing divertor target plates and the divertor dome. The reflectance of the surface of the tungsten divertor block is 23% at Hα (λ656.3 nm). A sandblast-processed tungsten surface effectively reduces direct reflectance; the resultant reflectance is less than 0.7%.

The optical design of Impurity Influx Monitor (divertor) was carried out for new ITER design and ray-tracing analysis shows that the spatial resolution of ITER requirement (50 mm) will be achieved by these optical systems designed here. The mechanical design of front end optics also carried out based on the optical design and results of port integration. In the upper port, front end optics can be installed inside the pipe of inner diameter of φ300 mm. In the equatorial port, all the optical component are placed to avoid the interaction with other diagnostic equipments. Heat analysis was carried out for the optimization the cooling method of mirrors, mirror holders and mounting modules to reduce the temperature rise caused by the nuclear heating. It indicates that mirrors can be cooled by the thermal conduction using mirror holders made of a copper alloy and making many cooling channels on the mounting module. Finally, the effect of the thermal deformation to the optical properties is estimated by using the optical design code.

Detailed spectroscopic studies of impurities have been made on the hydrogen plasma in the wavelength region from 10 to 1300 A with the upgraded DIVA tokamak device with a divertor. The toroidal field is 20 kG and the peak plasma current is 40 kA. Oxygen is the dominant impurity with the density of about 2% of the electron density. The effective ionic charge estimated from the amount of low-Z impurities (oxygen and carbon) is 2.4 to 3 in agreement with the value derived from the plasma conductivity. The emission spectrum of many lines observed in the region from 42 to 200 A is due to gold as impurity. Its spatial distribution has a peaked profile at a shorter wavelength and a hollow profile at a longer wavelength. The radiation power losses due to low-Z impurities and gold are estimated to be 30–35 kW and 27 kW, respectively, and the total power loss corresponds to 41–44% of the Joule-input power.

The impurity transport characteristics of ohmically and neutral beam heated diverted plasmas in JT-60 have been studied. In sawtoothfree, ohmically heated plasmas with Ip = 1.5 MA, e = (2−3) × 1019 m−3, q(a) = 3.2, no radiation buildup is observed. The temporal evolution of injected titanium has been measured by VUV and X-ray crystal spectrometers and PIN diodes. The spectral lines of Ti XX (259.3 A) and Ti XXI (2.6097 A) are compared with one-dimensional diffusive/convective impurity transport calculations. The transport calculations for Ohmic discharges with Ip = 1.5−2 MA, e = (1–3) × 1019 m–3, q(a) = 3−3.7, show that DA =(1−2) m2⋅s−1 and that the inward convective velocity is small. In neutral beam heated plasmas with Ip = 1.5 MA, e = 4 × 1019 m–3, q(a)= 3.6, Pb = 8 MW, no difference is observed in the global time evolution of the Ti XXI line emissions between co- and counter-injected discharges. The experimental observations are well represented by choosing DA = (1–2) m2⋅s−1 and a small inward velocity.

Divertor actions for controlling the impurity are investigated in DIVA, The heat exhaust function reduces the electron temperature of the plasma which contacting to the first wall, thereby reduces the ion-sputtering yield. Cooling of the scrape-off plasma with gas feeding further advances this situation and suppresses the contamination by the high- Z impurity. For a carbon and aluminum injection, the accumulation of impurities in the main plasma is reduced by the factor of 2–4 by the shielding action. The shielded impurities are rapidly guided into the burial chamber with a flow velocity roughly equal to that of the bulk scrape-off plasma. The back-flow of impurity ions from the burial chamber to the main plasma is verified to be small.

A pulse of methane gas and a short burst of aluminium have been injected into DIVA plasma at the point farthest from the divertor. For the same amount of injected impurity, the accumulation of injected impurity in the main plasma is reduced by a factor of 2 to 4 by the action of the divertor. The impurity shielded by the scrape-off plasma is rapidly guided into the burial chamber. The time needed for the poloidal excursion of the injected impurity ions in the scrape-off layer is roughly equal to that of the scrape-off plasma. Sweeping-out of the impurity ions diffusing from the main plasma into the burial chamber is also observed.

The many closely crowded lines (pseudo-continuum) of high-Z gold impurities are studied in detail with a calibrated grazing-incidence monochromator in hydrogen plasma, in the upgraded DIVA tokamak device with a divertor; the toroidal field is 20 kG and the peak plasma current is 40 kA. The emission spectrum of the many lines is observed in the wavelength region of 42 to 200 A. The time-resolved spatial distribution of the emission of the many lines was measured in each element in the wavelength region. It was found that the radiation due to the many lines in the wavelength region of 120 to 200 A concentrated in an intermediate region of the plasma column, and that the emission of the many lines in the wavelength region of 42–110 A peaked at the plasma centre. The radiation power loss due to the many lines in the 120–200 A region was dominant during the whole discharge. The radiation power due to the many lines corresponded to 40–70% of the total radiation power loss from the plasma volume.

Plasma-wall interactions in DIVA are described. Plasma particles are reflected at the surface of the copper shell covered with gold films (direct recycling). Distibution function of electrons in the scrape-off layer is not a simple Maxwellian, and has a high energy tail. The high energy component of electrons causes a relatively large heat transmission to the floating plate. The divertor reduces the contamination of the plasma due to heavy impurities, which irradiate most of the radiation loss from the plasma. Preliminary experiment about shielding effect of the scrape-off layer with respect to a methane gas injection are performed. Shielding effect is estimated to about 60 %.

We obtained clear evidence of metal impurity behaviour in DIVA by using a blow-off technique. The impurity in plasma was found to have a constant velocity of (2–3)·103 cm/s which is much less than the injected velocity of 105 cm/s. Confinement times of metal impurity of about 1 and 4 ms in the peripheral and central region respectively were obtained. The particle balance is represented as the reflected rate 78%, the reducted rate by the divertor action 16%, and the accumulated rate 6%. Major disruption was forced by injecting a large amount of gold atoms.