Graphite Tiles Research Articles

Preload Analysis of Screw Bolt Joints on the First Wall Graphite Tiles in EAST

The first wall in the Experimental Advanced Superconducting Tokamak (EAST) used graphite tiles to withstand high thermal energy. The graphite tiles are mounted on the heat sink using screw bolts which have been preloaded to produce a clamp force. The clamp force is very important to keep the graphite tiles tightly on the surface of the heat sink so that the heat flux crosses this contacting surface in a small thermal resistor. Without the clamp force, the small gap between the graphite tiles and the heat sink will make it impossible for thermal power to be carried away by cooling water. Some bolts may even fall off with the loss of clamp force. From the mathematical models, the loss process of the clamp force has been studied. Research results explain how the different thermal expansions of three members of the screw joint makes the clamp force decrease to zero under temperature rise and external force, and how the stiffness affects the relation between the clamp force and temperature. The research also gives the critical temperature at which the clamp force can remain above zero. Analysis results indicate that the current screw joints are almost destined to lose their clamp force during the running time of EAST, so the bolt joints should be redesigned in order to improve its reliability.

Damage in graphite tiles irradiated with helium plasmas

Plasma-facing materials in fusion devices are irradiated by a high particle flux from a dense plasma. Graphite tiles, widely used as plasma-facing materials, were irradiated with a helium plasma of a high flux and low energy; the damage in the graphite-tile target was investigated. Changes in the surface morphology were observed from scanning electron microscopy; the electrical conductivity was decreased from four-probe measurement; the intensity of the graphite disorder peak was increased from Raman spectroscopy.

R&D and maintenance on graphite tile of divertor region at EAST

EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45MPa. In the other hand, the structure of the graphite tiles is optimized.

Analysis of Limiters for ADITYA Tokamak

ADITYA Tokamak is a medium size ohmically heated tokamak. The hot plasma interacts with the vacuum vessel wall introducing impurities in the plasma and also damages the vacuum vessel wall. Limiter is used to reduce plasma - wall interaction and protect the vacuum vessel wall as well as inside components. It also controls impurity generation. ADITYA has a set of two types of limiters (i) Safety limiter (ii) a Poloidal limiter. Formally all the limiters are made of shaped graphite tiles fixed on stainless steel base plates. But graphite tiles are responsible for low Z impurities in plasma. To avoid these low Z impurities Molybdenum tiles are suggested. In this paper, we are going to present ansys analysis of surface temperature rise of molybdenum tiles compared to graphite tiles when used as Poloidal Limiter during plasma discharges in Aditya tokamak. The analysis is carried by ANSYS 11.0 software using Transient Thermal Analysis module. The results are compared with experimental results for graphite tiles. The graphite and molybdenum tiles are analyzed under same conditions.

SIMS analysis of the interaction between plasmas and the graphite first wall in RFX‐mod

Modified Reversed‐Field eXperiment is an experiment of magnetic confinement of thermonuclear plasma in the Reversed‐Field Pinch configuration. The toroidal device is designed to operate with hydrogen plasmas, and the most internal wall is entirely covered with polycrystalline graphite tiles. The disadvantage of graphite walls is the progressive hydrogen accumulation in graphite during operation. To overcome this problem, some graphite conditioning procedures are necessary (e.g. boronization, lithization and glow discharge). The critical role of graphite walls on plasma control suggested the use of secondary ion mass spectrometry (SIMS) technique for the evaluation of conditioning procedures efficacy. SIMS analyses of graphite samples after their exposure to plasmas or wall conditioning treatments are presented in this work. The SIMS technique resulted fundamental on the investigation of the boron distribution in the graphite tiles, and sample comparisons indicated the need for glow discharge cleaning system upgrading to improve the conditioning process and boronization uniformity. Copyright © 2012 John Wiley & Sons, Ltd.

Studies on scattering of laser radiation from viewing dump in tokamak Thomson scattering system

Thomson scattering is among one of the primary diagnostics used for tokamak plasma temperature and density measurements. Spuriously scattered stray signal further deteriorates the already existing poor signal-to-noise ratio in this technique. The present paper reports experimental investigations on the angular distribution of the scattered/stray radiation from graphite and a stack of laser blades dumps for different angles of incidence. Experimental results show a dependence of the stray signal on the dump-edge orientation. It is found that horizontally oriented dump edge produces a minimum level of stray light (around two orders less than in case of plane graphite tiles) in the horizontal plane, in which detectors for most of the diagnostics are installed. Studies are also made for p-polarized and s-polarized incident laser beams of 632.8 nm and 1064 nm wavelengths.

Tritium uptake in graphite tiles exposed to EAST plasma and then tritium gas

Tritium exposure experiments were carried out for three kinds of EAST SiC coated doped-graphite (SiC/C) samples, one from the original graphite tiles without being irradiated, and the other two from erosion and deposition areas of first wall after the 2009 campaign in EAST. β-ray-induced X-ray spectrometry (BIXS) was used to characterize the exposed samples. It is showed that the significant amount of tritium was absorbed in the surface of deposition sample in comparison with that of original sample, which was also supported by the results of imaging plate (IP) measurements. In addition, it was found that drastic decrease in tritium retention appeared by lowering exposure temperature, and the trapped tritium was maintained stably with time. Computer simulation is used to analyze the details of depth profile of tritium in different kinds of samples.

Key Features in the Operation of KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is working toward AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn, which provide high TFs up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma-facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350°C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor jacket of the magnet coils as well as a nonferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and in-vessel control coils (IVCCs) are key components to realize AT operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010, and diverted plasmas with high elongation (κ~1.8, δ>;0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned for 10 s. Finally, ELMy H-modes were successfully produced with 1.3-MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and was done at B <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">T</sub> = 2 T with I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">p</sub> = 0.6 MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the international thermonuclear experimental reactor (ITER) and future fusion reactors.

RFX-mod wall conditioning by lithium pellet injection

Plasma–wall interaction is one of the most important issues that present magnetic confinement devices have to face. In the RFX-mod reversed field pinch experiment plasma–wall interaction has become a hard point increasing plasma current up to the RFX-mod maximum design value of 2 MA, since in this case local power deposition can be as high as 10 MW m−2. Since the first wall of RFX-mod is entirely covered by graphite tiles different techniques have been tested to control hydrogen wall influx: He glow discharges cleaning, He discharges at high plasma currents, wall boronization and baking. With the best results obtained by boronization, at high plasma currents all such techniques improve the situation but do not allow a complete and stationary hydrogen influx reduction. Furthermore, in the presence of localized high power load the wall still responds providing very high influxes. In order to improve this situation wall conditioning by lithium has been tested. As a first lithization method to deposit a controllable amount of lithium on the wall, a room temperature pellet injector has been used (maximum pellet diameter of 1.8 mm and maximum length of 5 mm). Lithium coatings with a theoretical thickness of about 10 nm have been applied both to clean graphite tiles and over boronized ones. Lithization demonstrated to be effective in lowering hydrogen wall recycling to a value smaller than that of boronized graphite, with the effect lasting 20–30% more than in the boronized case. Compared with boronization, lithization slightly improves (by about 30%) particle confinement time and also clearly affects edge particle transport providing a lower edge density and more peaked density profiles. Lithization also reduces carbon content by about 10% over boronization but still no clear improvement has been observed in terms of energy confinement. Similar results have been obtained performing lithization over boronized graphite.

Spectroscopic study of plasma-facing graphite tiles modified by electron and proton irradiation

We have carried out Raman, electron spin resonance (ESR), and X-ray photoemission (XPS) spectroscopy measurements on the graphite tiles modified by electron and proton irradiation. While increase in local order was observed in the electron-irradiated sample, structural disorder and amorphization were revealed in the proton-irradiated sample, with much reduced electrical conductivity. Besides, a narrow ESR peak attributed to localized spins, closely associated with C–OH bonds as identified by the XPS measurements, was detected in the proton-irradiated sample.

Fuel re-absorption by thermally treated co-deposited carbon layers

Systematic studies have been conducted to address the fuel re-absorption by carbon deposits under repeated exposure to plasma after cleaning procedures. The investigation was done with graphite tiles from ALT-II (Advanced Limiter Test II), i.e. the main limiter at the TEXTOR tokamak. Pure graphite plates were used as the reference material. The experimental programme comprised the following: pre-characterization of specimens; D desorption by baking the tile at 1273 K; surface analyses of the fuel-depleted layers; exposure to deuterium in a laboratory plasma device and in TEXTOR; and quantitative assessment of deuterium re-absorption. The main result is that fuel retention in the re-exposed deposits is 30–40 times lower than that in the original co-deposit, showing that fuel re-absorption does not lead to an immediate re-saturation of deposits. Annealing at high temperatures enhances layer brittleness, leading eventually to detachment of co-deposits.

Advanced x-ray imaging of metal-coated/impregnated plasma-facing composite materials

A combination of x-ray imaging techniques is employed in the characterization of the coated/impregnated carbon-based composite materials to be used as plasma-facing components (PFC) in fusion devices. X-ray micro-tomography (μXCT) is applied in the visualization of carbon fiber composites (CFC)–Cu joined samples and in the characterization of CFC material porosity networks. Quantitative determination of the thickness of tungsten coating on carbon materials is performed using a combined absorption/fluorescence x-ray technique. The method was applied in the analysis of W-coated fine-grain graphite and CFC tiles. The x-ray imaging techniques allow rapid analysis with high spatial resolution.

Tritium Distribution on First Wall Carbon Tiles in JT-60U

T retention and its depth profile in the graphite tiles used for first wall of JT-60U have been measured by a tritium imaging plate technique and a full combustion method. T was found only limited depth beneath the plasma facing surface and little in both the surface region shallow than 1 μm and in bulk more than 1mm in depth. Although most of T produced by DD reactions are thermalized and neutralized in plasma and impinge on the plasma facing surface and penetrate into the inner surface, they are isotopically replaced by subsequently incoming D. Only some of high energy T escaping from plasma are directly implanted beneath the surface and retained escaping from the isotopic replacement until attainment of a saturation concentration.

Pulsed repetition rate nanosecond laser heating and ablation of the tokamak graphite tile deposited layers

Laser heating and ablation of the plasma-facing surface of a graphite tile from TEXTOR tokamak that was covered by a deposited carbon layer has been studied. Laser heating measurements were performed with a pulsed nanosecond Nd-YAG laser (2nd harmonic, 10 kHz repetition rate, 100 ns pulse duration). Surface temperature measurements were recorded with a home-made pyrometer having a response time of 15 μs ( t 99%). The experimental results are simulated with an analytical model of laser heating of a surface covered by a deposited layer and heated repeatedly by laser pulses. The comparison between experimental and theoretical data of the observed temperature excursions enables us to assess the deposited carbon layer physical parameters (thermal conductivity, porosity, etc.) if the thermal and optical properties of the graphite substrate are known. Laser ablation measurements were performed with two pulsed nanosecond Nd-YAG lasers (20 Hz and 10 kHz repetition rate with 5 ns and 100 ns pulse duration, respectively). For a plasma-facing graphite surface covered by a thick (∼30–50 μm) deposited carbon layer, the ablation threshold is 4.5 ± 1 kJ/m 2 regardless of the pulse duration. The obtained ablation threshold is significantly lower than the one measured for a virgin tokamak graphite sample. The comparison of the experimental results and theoretical data demonstrated that the laser ablation mechanisms for tokamak graphite and thick carbon layers deposited on plasma-facing surface are different.

Neutral Beam Armor for NSTX Upgrade

The National Spherical Torus Experiment (NSTX) is a low aspect ratio, spherical torus (ST) configuration device which is located at Princeton Plasma Physics Laboratory (PPPL). This device is presently being upgraded to enhance its operations by adding a second Neutral Beamline (NBL). This change will nearly double the power available to the plasma but necessitate improvements to other design aspects of NSTX. Included in these upgrades are the relocation and upgrade of the NSTX Neutral Beam Armor to capture both sets of beamline source profiles while maintaining the same level of vacuum vessel wall protection.In order to minimize the space required to accomplish this, it has been proposed to relocate and reuse the existing armor array, improving the design so that two overlapping sets of beam profiles both fit completely. This beamline overlap could possibly cause the armor tiles to experience higher heat fluxes which translate into higher internal mechanical stresses. This would be mitigated by changing the isotropic graphite (ATJ) tiles in the overlap areas to a rugged 3D carbon-fiber composite (CFC) material, capable of handling thermally-induced stresses. Additional benefits to this recycling design proposal include opportunities to reduce project cost, increase diagnostic port access, and improve an awkward and difficult mounting scheme.

The design and application of the movable limiter in EAST

There are two movable limiters that have been installed on EAST. The movable limiter consists of three parts that are blade, support components and drive components. The blade which consists of heat sink and graphite tiles has to face plasma during plasma discharge. The support components consist of a long shaft and a movable plate. The drive components consists of a screw shaft, a step motor, four guide shafts and two support plates. There is an extra support which has fit in Tokamak and a frame on the outside of the flange. The specific shape of the blade has been optimized so that the graphite tiles on the blade can withstand max 5 MW/m^2 heat load with two long cooling channels. The long shaft of the transfer components connect with blade to make it can move along the axis of the limiter. The movable blade could adjust a wide variety of plasma major and minor diameters. The drive components provide motion from step motor which can be remote controlled in the control room far away from the scene. The rotation is translated to slide motion by a screw shaft. The screw shaft can push the slide which connected with the long shaft to at last move the blade. To introduce the motion from outside to the vacuum vessel, the limiter has a long bellow and special structure as dynamical seal. This seal structure proved very reliable. The remote control system has been also equipped the distance sensor to provide the information of the position of the limiter. The control system is based on PC so it can be easily operated. The two movable limiters have been installed on the EAST. During past twice discharges, the movable limiter proved to be a useful tool for EAST to control the plasma boundary.

Deuterium retention in SiC coated graphite by D 2+ implantation

The doped graphite tiles bolted to the active cooling heat sink, made of GBST1308 (1% B 4C, 2.5% Si, 7.5% Ti) coated with SiC, are now being used as the only plasma facing material (PFM) for the EAST device since the campaign of 2008. From the plasma density and fueling point of view, it is important to study thoroughly the hydrogen isotope retention in this kind of SiC-coated doped graphite. D 2 + implantations into the SiC coated doped graphite were performed at Shizuoka University. The chemical states of Si and C were studied by means of X-ray photoelectron spectroscopy (XPS), and the thermal desorption behavior of deuterium was analyzed by thermal desorption spectroscopy (TDS). It was found that deuterium was trapped by both C and Si in the SiC coatings. In the previous studies, Oya et al. reported the deuterium retention behavior in polycrystalline β-SiC. In this paper, difference of retention behavior in β-SiC and SiC coating will be also discussed.

The loosing mechanism of screw bolts on the first wall graphite tiles in EAST

The first wall in Experimental Advanced Superconducting Tokamak (EAST) has used many graphite tiles to face the plasma. All graphite ties have been mounted on heat sink using screw bolts which had been preloaded to produce clamp force. The clamp force is very important to keep the graphite tiles on the surface of the heat sink tightly because the heat flux will cross this contacting surface in a small thermal resistor. Without the clamp force the small gap between graphite tiles and heat sink will lead thermal power cannot be carried away by cooling water. The worse is some bolts turned out with the loss of clamp force. From the mathematical models the process of the loss of clamp force has been studied. Research results explain how the different thermal expansion of three members of the screw joint leads to the clamp force decrease to zero under temperature rise and external force and how the stiffness affect the relationship between the clamp force and temperature. The research also gets the critical temperature point that the clamp force can remain above zero. Research results indicated this screw joint is absolute to lose their clamp force during the EAST running so that the bolt joints should be redesign to improve its reliability.

X-ray microbeam transmission/fluorescence method for non-destructive characterization of tungsten coated carbon materials

Tungsten erosion, its subsequent transport and redeposition are of great interest, because a full tungsten divertor is foreseen to be used during the deuterium–tritium operational phase of ITER. The erosion of tungsten and carbon marker layers was extensively studied in the outer divertor of ASDEX Upgrade (AUG), and work is currently in progress to completely replace the existing JET CFC tiles with tungsten-coated tiles within the JET ITER-like wall project. The need for fast and non-destructive method which allows the quantitative determination of the thickness of a tungsten coating on a carbon material on large areas led us to evaluate a combined absorption/fluorescence X-ray (XRTF) technique. The method can provide fast analysis, high spatial resolution and a material selective detection of deposited layers and inclusions. It was applied on W coated fine grain graphite (FGG) tiles from AUG's divertor. It is proved that the method is able to provide information about the uniformity of the tungsten coating on a graphite or CFC substrate whilst the technique can be used to determine the thickness of the tungsten and other marker materials coatings. It represents a unique instrument for the post-mortem analysis of the coatings.

Liquid Lithium Limiter on the RFX-mod experiment

In this work, we present the experimental apparatus of the Liquid Lithium Limiter that will be used for conditioning the first wall of the RFX-mod experiment. The system is based on a Capillary Porous System, on loan from ENEA-Frascati laboratories, where it has been successfully exploited for similar purposes on the FTU Tokamak. In RFX-mod, the limiter position and temperature are remotely controlled; a near-infrared detector is used to measure the surface temperature during plasma exposures. A set of graphite samples for post-mortem analysis and a quartz micro-balance system are used to monitor the lithium deposition on the graphite tiles completely covering the inner wall. Wall treatment conditioning procedures will be investigated and compared exploiting the RFX-mod flexibility of running as a reversed field pinch or a Tokamak.