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A new temperature measurement method—the dual two-color method—was developed to accurately measure the temperature over an ultra-wide temperature range (200–3600 °C) for ITER divertor infrared thermography. This novel method introduces a third wavelength filter to the conventional two-color method by replacing the shorter single wavelength bandpass filter with a customized dual-bandpass filter having two transmission bands, without having to add a third infrared camera. The dominant wavelength band of the total radiance through the dual-band filter changes automatically as the temperature increases and, consequently, either the shorter or longer wavelength band of the dual-bandpass filter is used to establish the two-color combination at both low and high temperatures. The dual two-color method increased the acceptable measurement error of the two-color radiance ratio for the temperature measurement requirement of the ITER divertor infrared thermography to 9.45% from that of 4.3% when using the conventional two-color method.

ITERダイバータ赤外サーモグラフィでは高い放射線環境下にミラー及びレンズを配置し光を伝送する。そのため、ITERでの全運転期間を通して顕著な光学特性の劣化がない光学材料の選定及び遮蔽設計が必要不可欠となる。本研究ではITERダイバータ赤外サーモグラフィで使用可能な光学素子の選定及びその遮蔽設計の指標の決定を目的とし、レンズ材料のガンマ線及び中性子照射を実施しその前後の光学特性の計測から中赤外領域における放射線耐性を評価した。今回はその中でもSi、ZnSe、ZnS、CaF2の主要材料に絞り各材料のガンマ線及び中性子照射特性の詳細を報告する。, 第38回プラズマ核融合学会・年会

本研究ではITERダイバータ領域用の紫外-近赤外域分光計測システム開発の一環として、タングステンダイバータ化したITERにおける燃料/壁材由来の中性粒子及びイオンの発光とそれに寄与する素過程について検討を行っている。現在設計中の分光システムで計測され得る各種発光強度分布に加え、これらを用いてイオン化フロント位置や粒子発生量等の物理量を評価するために重要な素過程であるイオン化/再結合の寄与等について、ITERのプラズマシミュレーションデータから計算した結果を報告した。, 第38回 プラズマ・核融合学会 年会

ITERではタングステン(W)ダイバータ板等で反射した光の混入によるプラズマ発光計測の精度悪化が問題視されており、この反射特性は壁面の表面状態に依存する。本研究では、様々な熱負荷を与えたITERダイバータ用Wモノブロックに対してゴニオフォトメーターを用いて双方向反射率分布関数(BRDF)を計測した。熱負荷を与える前のWモノブロックでは切削痕によって異方的な反射特性を示す一方、熱負荷を与えたWモノブロックでは再結晶化によって表面が粗くなり、拡散的な反射特性を示した。熱負荷による表面構造変化に応じたBRDFの変化について報告する。, 第38回 プラズマ・核融合学会 年会

ITERのダイバータは光の反射率の高いタングステン（W）で製造されるため、プラズマ分光計測器の測定光にダイバータ表面からの反射光が混入し、プラズマ発光量評価の精度が低下する恐れがある。反射光はダイバータ板表面の状態によって特性が変化するため、光の反射特性とダイバータ表面の状態の関係を把握することが重要となる。 本研究では、直線型プラズマ装置NAGDIS-IIで生成したアルゴンプラズマの照射によりスパッタリングしたW試料の双方向反射率分布関数（BRDF）をゴニオフォトメータを用いて調べた。プラズマ照射前のW試料表面には紙やすりで傷を作成し、ITERダイバータのWモノブロックの表面粗さ（Ra）を模擬している。W試料へのイオン入射エネルギーを変化させたときの、RaとBRDFの変化について報告した。, The 30th International Toki Conference on Plasma and Fusion Research

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.

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.

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.

Japan Domestic Agency (JADA) is responsible for six diagnostic systems in the ITER project. We have successfully developed a prototype instrumentation and control (I&C) system for the ITER divertor thermocouples system. The prototype I&C system includes a supervisory function, a sequencing management function, and a data acquisition function. The supervisory function, which was implemented using EPICS (Experimental Physics and Industrial Control System), manages internal operations for measurement such as health check of sensors. We developed a conversion tool from flowcharts to EPICS record source codes. The EPICS records generated by the conversion tool are used to trigger each operation step and to indicate the progress of the sequence. We also developed the sequencing management function that coordinates the execution of operation steps. Both the flowchart conversion tool and sequencing management function were developed to reduce designers’ mistakes and to ensure rapid development. The prototype I&C system for the DTC was tested. We confirmed that the I&C system performed satisfying the requirements. This validated design will be applied to other ITER diagnostic systems procured by JADA.

In ITER, it is important how the CODAC system conducts many plant systems including diagnostic systems. In order to establish necessary communications between the diagnostics systems and the CODAC system, Japan domestic agency (JADA) has proposed the new concept of supervisory system for the diagnostic system based on our experiences in operating plasma diagnostic systems. The supervisory system manages operation sequences, current state and configuration parameters for the measurement. JADA designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into the EPICS (experimental physics and industrial control system) records source codes is introduced. This tool will ensure reduction of the system designers’ efforts. We designed a communication protocol to configure measurement parameters and proposed configuration parameter validation function. We also analyzed the management of the central/local control mode for the diagnostic systems. The function which selects the adequate limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is proposed. JADA will develop a prototype of the supervisory system and validate the design in 2013.

A prototype of the ITER pulse scheduling system that prepares and manages parameters for ITER plasma operations has been designed. Based on the analyzed requirements on the system, structure of the parameters and necessary functions were determined. Segment and module structures were tuned to the ITER requirements. Three types of validations assure sanity of the parameters. The design limits check and the operation window check verify whether the values of the parameters do not exceed the limits. The consistency check calculates dependency among parameters in accordance with logics described in a scripting language. The ITER pulse scheduling system provides interface with a physics model and simulator. Some abstract physics parameters are converted to engineering parameters with the physics simulation. The results of simulation such as plasma characteristics of specified parameters are also shown to the researchers. The tool to specify the parameters is data-driven. Therefore, it is flexible for changes of number of the parameters. A prototype is being implemented in 2011. Using the prototype, this design will be verified and refined. The evaluation of the prototype will be a basis of the final production of the ITER pulse scheduling system.

Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

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.

In JT-60U, Xe gas has been injected into reversed shear plasmas with internal transport barriers. During these discharges, spectra from highly ionized Xe atoms have been observed in a wavelength range of 4–16 nm. The observed spectra have been analyzed by calculating theoretical spectra for Xe XXXVII–XLV 3s–3p and 3p–3d transitions with relativistic atomic structure codes, a collisional radiative model, and an impurity transport code. Several lines for Xe XLI and XLII have been newly identified. Soft X-ray emission profiles have also been analyzed, and the results showed that the Xe ions accumulated inside the internal transport barrier.

Quartz glasses of KU-1 and KS-4V are candidate window materials for the ITER optical diagnostics in UV and visible range. The temperature dependence of the radiation-induced transmission losses in UV range has been investigated under γ-ray irradiation for the KU-1 and KS-4V quartz glasses up to 10 MGy. The KU-1 and KS-4V quartz glass samples were irradiated at room temperature, 100, 200 and 300 °C. Dose rate was 0.46–1.1 MGy, and 2.6 Gy/s from 1.1 to 10.4 MGy. It was confirmed that there is no significant loss in wavelength range longer than 350 nm for both window materials. Transmission loss in KU-1 is larger than that in KS-4V under temperature below 200 °C. KU-1 has large temperature dependence. On the other hand, temperature dependence is not clear in KS-4V above 100 °C. Prominent recover of the transmission loss was not observed after irradiation for each sample. From the transmission loss point of view, KS-4V is better window material than KU-1 at temperature below 200 °C. KU-1 is available at 300 °C.

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.

This report summarizes the results of in situ measurements of noninductive voltages developed across prototype International Thermonuclear Experimental Reactor (ITER) magnetic coils performed at the Japan Materials Test Reactor. The voltages appear to be mainly thermoelectric in origin, with the thermal sensitivity developing gradually through irradiation and the thermal gradients supplied by nuclear heating. Possible mechanisms and circuits within the coils that can contribute to this radiation-induced thermoelectric sensitivity are described, and schemes to improve the ITER coil design are outlined.

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.

Metallic materials such as molybdenum and tungsten will be used for plasma facing mirrors in ITER. Test pieces of molybdenum and tungsten mirrors manufactured by different methods were irradiated by high flux (5.2–6.0 × 10 19 m −2 s −1 ), low energy (67–80 eV) deuterium ions. The molybdenum mirror manufactured by sintering and electron beam melting of a thin layer of the surface had a higher reflectivity before irradiation than the mirror made without surface melting, but the reflectivity of the former mirror decreased rapidly with fluence. This rapid decrease was caused by blisters generated on the mirror surfaces. The decrease of the 250 nm–20 μm wavelength reflectivity of the molybdenum and the tungsten mirrors made by sintering without surface melting was less than 10% at fluence of 1.2–1.3 × 10 25 m −2 .

Magnetic coils are indispensable for plasma position control in tokamaks. For ITER, non-inductive voltages generated under irradiation have to be reduced to well below 1 μV in order to support long pulse (3000 s) operation. In situ measurements of the differential voltage between the ends of two magnetic coils wound with mineral-insulated (MI) cable have been carried out at the JMTR fission reactor. The MI cables of the two magnetic coils had copper center conductor with diameter 0.5 and 0.8 mm, respectively, stainless steel outer sheath and MgO insulation. The measured differential voltage for the two MI cables increased with neutron fluence, reaching 4.5 and −0.7 μV in the two coils, at a fast neutron fluence of 1.26 × 1023 n/m2. The magnitude of the measured voltage can be explained by thermoelectric potentials, enabled mainly by non-uniform transmutation and displacement damages of the copper core of the cable.

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.

The transmissivity of KU-1 quartz which is a candidate of the window material for UV and visible spectroscopy in ITER was measured under the irradiation of 14 MeV neutrons up to the fluence of 7.4×1019 n/m2 in the UV range (200–400 nm). Significant transmission loss was observed in the wavelength range of 200–300 nm. Two absorption peaks were identified: an E′-center at 215 nm and Si(III) defects at 245 nm. The E′-center absorption is much larger than the Si(III) defect absorption. The transmission loss increased with the neutron fluence, decreased during the irradiation breaks and returned almost to the same level as before when the irradiation resumed. This phenomenon indicates that an absorption induced by the neutron irradiation is caused by stable scattering and short lifetime centers.

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 diagnostic system of JT-60U (JT-60upgrade) is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters have been improved, so ...

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has b...

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.

Irradiation tests on many diagnostic components have been carried out as a part of the R&D program of the ITER Engineering Design Activities (EDA). Five kinds of ITER round robin fibers were irradiated in JMTR and60Co γ-rays. Induced transmission loss of those fibers are much smaller than that of pure SiO2 core fiber. Especially, KU-H2G from Russian Federation and F-doped fibers from Japan have rather good radiation hardness even in the visible range. Those fibers might be available in the diagnostic port of ITER. The transmissivity of KU-1 quartz, which is a candidate of the window material for UV and visible spectroscopy in ITER, was measured under the irradiation of 14 MeV neutrons and γ-rays in the UV range (200 - 400 nm). Significant transmission loss was observed in the wavelength range of 200-300 nm. The Mica substrate bolometer was irradiated in JMTR during three irradiation cycles. Total neutron fluence was about 1×1024 n/m2 which was equivalent to 0.1 dpa. Significant increase in the meander resistance was observed, which was causedby the nuclear transmutation of gold into mercury. The use of gold meanders might be problematic in ITER.

As a part of the R&D program of the ITER Engineering Design Activities (EDA), we have carried out irradiation tests on a number of diagnostic components. In this paper the recent results of neutron irradiation tests on diagnostic windows, bolometer substrates and magnetic coils at JAERI are described. UV range transmission losses of a KU-1 quartz glass were measured during 14 MeV neutron irradiation at the Fusion Neutronics Source (FNS) with neutron fluences of up to 7.4×1019 n/m2. Significant transmission losses were observed in the wavelength of 200–300 nm. Mica substrates of bolometers were irradiated at the Japan Material Testing Reactor (JMTR) with neutron fluence between 1×1022 and 2.2×1023 n/m2. The size expansion was found to be only 0.13% at 2.2×1023 n/m2 and 150 °C, which gives us the good prospect of the bolometer in the ITER-FEAT application. Magnetic coils made with four kinds of Mineral Insulated (MI) cables have been irradiated in JMTR up to a fast neutron fluence of 1.1×1024 n/m2. Radiation Induced Electrical Motive Force (RIEMF) of several volts was observed between core inductor and sheath of the MI cable.

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.

In order to obtain improved confinement plasmas with high radiation at high density, Ar gas was injected into ELMy H mode plasmas in JT-60U. A confinement improvement of HH98(y,2) ≈ 1 was obtained with a high radiation loss power fraction (~80%) at an electron density of ~0.65nGW. The HH factor was about 50% higher than that in plasmas without Ar injection.

Recent development of the JT-60U charge exchange recombination spectroscopy (CXRS) diagnostic system is reported. The measurements of the radial profiles of the ion temperatures, the toroidal and poloidal rotation velocities, and the impurity density are based on CXR spectra of fully stripped carbon ions induced by neutral beam injection. Measurements are made simultaneously at 59 spatial points (23 toroidal and 36 poloidal) with a time resolution of 1/60 s throughout the full discharge pulse of 15 s. Considerable effort has been expended to ensure the automatic provision of reliable profile data, where the mechanical components of the system must cope with both a baked tokamak vessel (∼300 °C) and an expected disruption-induced acceleration of up to 60 g, where g denotes the gravitational acceleration, i.e., 9.8 m2/s.

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.

Collective Thomson scattering (CTS) diagnostic based on a pulsed CO 2 laser (wavelength 10.6 μm) has been developed to establish a diagnostic method of confined α-particles in burning plasmas. A high-repetition and high-energy transversely excited atmospheric (TEA) laser has been developed as a source of the CTS diagnostic. In order to obtain single-mode output, which is needed for CTS diagnostic, seed laser is injected into the cavity with unstable resonator. Pulse energy of 17 J with a repetition rate of 15 Hz has been achieved in a single-mode operation. This result gives a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires energy of 20 J with repetition rate of 40 Hz. Proof-of-principle test will be carried out in the JT-60U tokamak by using the newly developed laser. Preliminary consideration of the CTS diagnostic in the tandem mirror GAMMA 10 shows that axial profiles of ion temperature will be obtained using a circumferential collection mirror ofscattered power.

Beam stopping cross section and shine-through for neutral hydrogen beam injection into fusion plasmas have been calculated by using recommended cross sections presently available for atomic processes including multistep collision processes involving excited states. The shine-through thus obtained agrees well with recent experiments of JT-60U. The present calculations show that the multistep processes play a crucial role in the stopping of high-energy neutral hydrogen beams in high-density plasmas. Analytical fits to the stopping cross sections and fitting parameters are also presented for plasma impurities with nuclear charge and Z = 26.

As an R&D task of the ITER engineering design activities (EDA), we carried out the irradiation tests on the basic materials of transmission components such as ceramics, windows, fiber optics and mirrors, and in-vessel diagnostics sensors such as magnetic probes and bolometers. Ceramics will be used as general insulating materials. Here we investigated the radiation induced conductivity (RIC) of ceramics for 14 MeV neutrons. Optical elements are the most sensitive to radiation damages among the transmission components. Radiation induce transmission loss of window materials and fiber optics have been measured in the Japan material testing reactor (JMTR). Off line irradiation tests were carried out for molybdenum corner cube reflector (CCR) and gold coated mirrors in JMTR also. Fourteen MeV-neutron-induce luminescence of window material was evaluated in the fusion neutronics source (FNS). Bolometers should be installed inside the vacuum vessel to measure the radiation losses in the infra red to soft X-ray range. We demonstrated the performance of the JT 60 type bolometer under 60Co gamma-rays.