Divertor Tiles Research Articles

Investigation on the erosion/deposition processes in the ITER-like wall divertor at JET using glow discharge optical emission spectrometry technique

As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

Tritium distributions on tungsten and carbon tiles used in the JET divertor

Tritium distributions on the W-coated divertor tiles used with Be wall in JET 2011–2012 ITER-like wall (JET-ILW) campaign were measured using an imaging plate (IP) technique. The high intensity of photo-stimulated luminescence (PSL) from IP was observed at the regions covered by deposited Be layers. However, the PSL intensity was not simply proportional to the thickness of the deposited Be layers; the shadowed region of Tile 4 showed the highest PSL intensity though the thickness of deposited Be layer on this region was smaller than that on Tile 0 and the apron of Tile 1 by an order of magnitude. These observations indicated the influence of impurities such as oxygen on tritium retention in the deposited Be layers. The C tiles used in the 2007–2009 JET carbon wall (JET-C) campaign were also examined. The high PSL intensity was observed for the regions covered with deposited C layers in this case. The area of tile surfaces covered by the deposited tritium-rich layers on the W-coated-tiles used in the JET-ILW campaign was significantly smaller than that on the C tiles used in the JET-C campaign.

Two-dimensional numerical study of ELMs-induced erosion of tungsten divertor target tiles with different edge shapes

Thermal performance of the divertor tile with different edge shapes was assessed numerically along the poloidal direction by a two-dimensional heat conduction model with considering the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma. The energy flux density distribution arriving at the castellated divertor tile surface was evaluated by a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo Collisions code and then the obtained energy flux distribution was used as input for the heat conduction model. The simulation results showed that the divertor tiles with any edge shape of interest (rectangular edge, slanted edge, and rounded edge) would melt, especially, in the edge surface region of facing plasma poloidally under typical heat flux density of a transient event of type-I ELMs for ITER, deposition energy of 1MJ/m2 in a duration of 600μs. In comparison with uniform energy deposition, the vaporizing erosion was reduced greatly but the melting erosion was aggravated noticeably in the edge area of plasma facing diveror tile. Of three studied edge shapes, the simulation results indicated that the divertor plate with rounded edge was the most resistant to the thermal erosion.

Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak

The results of reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak by means of the EFIT code and by the method of movable filaments with the use of the data from magnetic measurements are compared. The EFIT code allows one to completely reconstruct the magnetic configuration by solving the Grad−Shafranov equation. In the method of movable filaments, the distribution of the toroidal current flowing through the plasma is described by a set of infinitely thin current-carrying rings. In this method, the last closed magnetic surface (LCMS) and the open surfaces lying beyond the LCMS are calculated. Using both methods, the coordinates of the regions where the separatrix strikes the divertor plates were determined. The results obtained agree well with the distributions of the temperature over the tungsten divertor tiles measured using an IR camera.

First dust study in JET with the ITER-like wall: sampling, analysis and classification

Results of the first dust survey in JET with the ITER-Like Wall (JET-ILW) are presented. The sampling was performed using adhesive stickers from the divertor tiles where the greatest material deposition was detected after the first JET-ILW campaign in 2011–2012. The emphasis was especially on sampling and analysis of metal particles (Be and W) with the aim to determine the composition, size, surface topography and internal dust structure using a large set of methods: high-resolution scanning and transmission electron microscopy, focused ion beam, electron diffraction and also wavelength and energy dispersive x-ray spectroscopy. The most important was the identification of beryllium dust both in the form of flakes and droplets with dimensions in the micrometer range. Tungsten, molybdenum, inconel constituents were identified along with many impurity species. The particles are categorised and the origin of the various constituents discussed.

Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24MW/m2 and absorbed energy between 370 and 680kJ. This loading results in maximum surface temperatures between 1300°C and 2800°C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results from the operation in the ASDEX Upgrade experiment are presented.

Development of laser induced breakdown spectroscopy for studying erosion, deposition, and fuel retention in ASDEX Upgrade

The paper deals with the development of laser induced breakdown spectroscopy (LIBS) into an in situ method for studying erosion/deposition processes at the first walls of fusion reactors. To this end, samples extracted from the divertor tiles of ASDEX Upgrade after the 2009 plasma operations were analyzed using LIBS for their composition and the results were compared with other post mortem deposition data. Quantitative depth profiles for the elemental concentrations were extracted from LIBS spectra by applying a novel data processing method. In addition, both multiline and multispot averaging procedures were applied to reduce fluctuations in the data. The LIBS concentration profiles matched qualitatively with those given by secondary ion mass spectrometry and quantitatively with the ion-beam data. The deuterium content of the samples could be reliably determined if the surface densities were >1017at/cm2.

Ammonia production in nitrogen seeded plasma discharges in ASDEX Upgrade

In present tokamaks nitrogen seeding is used to reduce the power load onto the divertor tiles. Some fraction of the seeded nitrogen reacts with hydrogen to form ammonia. The behaviour of ammonia in ASDEX Upgrade is studied by mass spectrometry. Injection without plasma shows strong absorption at the inner walls of the vessel and isotope exchange reactions. During nitrogen seeding in H-mode discharges the onset of a saturation of the nitrogen retention is observed. The residual gas consists of strongly deuterated methane and ammonia with almost equal amounts of deuterium and protium. This confirms the role of surface reactions in the ammonia formation. The results are consistent with findings in previous investigations. A numerical decomposition of mass spectra is under development and will be needed for quantitative evaluation of the results obtained.

Characterization of deuterium retention and co-deposition of fuel with lithium on the divertor tile of EAST using laser induced breakdown spectroscopy

A laser induced breakdown spectroscopy (LIBS) system has been developed to measure and monitor the composition evolution on plasma facing materials (PFMs) of Experimental Advanced Superconducting Tokamak (EAST). As a necessity and important proof of principle experiment, LIBS analysis has been performed for lithium–deuterium co-deposition layer diagnosis of EAST divertor tiles in lab experiments. The distribution of deuterium retention has been obtained from the depth of 0.5–4μm in the divertor tiles. The deuterium/hydrogen concentration ratio was estimated as 0.17±0.02 in lithium–deuterium co-deposition layer. Moreover, the depth profile behaviors of lithium and deuterium indicate that the deuterium retention in divertor tile came from lithium–deuterium co-deposition processes during deuterium discharge in EAST. This work would improve the understanding of deuterium retention and lithium–deuterium co-deposition mechanism and give a guidance to optimize the LIBS system which will be a unique and useful diagnostic approach in EAST 2014-campaign.

2D tritium distribution on tungsten tiles used in JET ITER-like wall project

Post-mortem measurements of 2-dimensional tritium (T) distribution using an imaging plate (IP) technique were performed for tungsten (W) divertor tiles (W-coated CFC) used in JET-ITER like wall (ILW) project. The observed T distributions were clearly inhomogeneous, and there were band-like regions with high T concentrations that extended in the toroidal direction on tiles 1, 3, 4 and 6. The concentrations of T in the band-like regions were higher by an order of magnitude than the concentrations in other parts. The inhomogeneous T distributions were explained by non-uniform co-deposition with other elements such as beryllium. The concentrations of T on the outboard vertical tiles (tiles 7 and 8) were low and relatively uniform in comparison with other tiles.

Impact of ELM filaments on divertor heat flux dynamics in NSTX

The ELM induced change in wetted area (Awet) and peak heat flux (qpeak) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger Awet and lower qpeak. The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n=1–5), while typical peeling–ballooning ELMs have higher mode number of n=10–20. For ELMs with smaller number of striations, relative Awet change is rather constant and qpeak change rapidly increases with increasing ELM size, while Awet change slightly increases leading to a weaker increase of qpeak change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

A reconsideration of the noise equivalent power and the data analysis procedure for the infrared imaging video bolometers.

The infrared imaging video bolometer (IRVB) used for measurement of the two-dimensional (2D) radiation profiles from the Large Helical Device has been significantly upgraded recently to improve its signal to noise ratio, sensitivity, and calibration, which ultimately provides quantitative measurements of the radiation from the plasma. The reliability of the quantified data needs to be established by various checks. The noise estimates also need to be revised and more realistic values need to be established. It is shown that the 2D heat diffusion equation can be used for estimating the power falling on the IRVB foil, even with a significant amount of spatial variation in the thermal diffusivity across the area of the platinum foil found experimentally during foil calibration. The equation for the noise equivalent power density (NEPD) is re-derived to include the errors in the measurement of the thermophysical and the optical properties of the IRVB foil. The theoretical value estimated using this newly derived equation matches closely, within 5.5%, with the mean experimental value. The change in the contribution of each error term of the NEPD equation with rising foil temperature is also studied and the blackbody term is found to dominate the other terms at elevated operating temperatures. The IRVB foil is also sensitive to the charge exchange (CX) neutrals escaping from the plasma. The CX neutral contribution is estimated to be marginally higher than the noise equivalent power (NEP) of the IRVB. It is also established that the radiation measured by the IRVB originates from the impurity line radiation from the plasma and not from the heated divertor tiles. The change in the power density due to noise reduction measures such as data smoothing and averaging is found to be comparable to the IRVB NEPD. The precautions that need to be considered during background subtraction are also discussed with experimental illustrations. Finally, the analysis algorithm with all the improvements is validated and found to reproduce the input power well within 10% accuracy. This article answers many fundamental questions relevant to the IRVB and illustrates the care to be exercised while processing the IRVB data.

Global helium particle balance in LHD

Global helium particle balance in long-pulse discharges is analyzed for the first time in the Large Helical Device (LHD) with the plasma-facing components of the first wall and the divertor tiles composed of stainless steel and carbon, respectively. During the 2-min discharge sustained by ion cyclotron resonance heating (ICRH) and electron cyclotron heating (ECH), helium is observed to be highly retained in the wall (regarded as both the first wall and the divertor tiles). Almost all (about 96%) puffed helium particles (1.3×1022 He) are absorbed in the wall near the end of the discharge. Even though a dynamic retention is eliminated, 56% is still absorbed. The analysis is also applied to longer pulse discharges over 40min by ICRH and ECH, indicating that the helium wall retention is dynamically changed in time. At the initial phase of the discharge, a mechanism for adsorbing helium other than dynamical retention is invoked.

Effect of steady magnetic field on laser-induced breakdown spectroscopic characterization of EAST-like wall materials

Our recent investigations were focused on the feasibility of the LIBS method proposed for EAST under vacuum conditions as well as with a magnetic field. Aluminum (replaced Be)–lithium alloys were used as a substitute for a uniform lithium deposition layer on the first wall. Detailed information of divertor tiles (multi-element doped graphite) and aluminum–lithium alloys were obtained by analyzing the spectra from 200 to 980nm. With the magnetic field (0.94T), various line emissions obtained from the constituents of samples shown an enhancement (>2times) in intensity due to the increase in the effective plasma density and temperature as a result of magnetic confinement. The effect of magnetic field on the emission intensity of LIBS at the different pressure (1.0×10−5–1000mbar) would help us to develop a quantitative LIBS approach to monitor impurity deposition and fuel retention on the first wall.

Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

Measurement of dust conversion factor for the JET carbon divertor phases

We compare the deposition of material on poloidal sets of divertor tiles that had been exposed in JET in 1998–2009 and 1998–2007. Post mortem analyses suggest toroidally integrated deposition being increased by 197.5cm3 during 2007–2009. The analysis of dust collected from the divertor indicates the amount accumulated during the same period to be 248.4g. Converting the weight of dust to volume, the fraction of material entering the divertor that was converted to dust and flakes is 43±10%. The size of most dust particles ranged from 10 to 100μm. The integrated amount of deposition on the “marker” tiles exposed in 2007–9 was found to be more than twice the amount expected from film growth on other tiles plus the dust because the plasma responds differently to the new tiles.

Analysis of rotating collectors from the private region of JET with carbon wall and metallic ITER-like wall

Rotating collectors are used in JET to provide time-resolved measurements of erosion and redeposition of vessel materials. The silicon collecting discs rotate behind an aperture, driven by pulsing of the toroidal magnetic field, with the deposits analysed ex-situ by nuclear reaction analysis. The angular dependence of deposition is mapped to discharge number using the B-field history, allowing the influence of different plasma configurations and parameters to be investigated. A simple geometrical model using sputtering and reflection from the strike point has qualitatively reproduced the deposition found on collectors located under the central divertor tile and facing towards the inner strike point. The beryllium deposition on the ITER-like wall (ILW) collector showed an order of magnitude reduction in deposition compared to carbon deposition on the JET-C collector. This decreased deposition is attributed to low long range divertor transport due to reduced chemical sputtering/erosion and codeposition of beryllium relative to carbon.

Comparison of tritium measurement techniques for a laser cleaned JET tile

Over the last 7–8 years, two quantitative analyzing methods—accelerator mass spectrometry (AMS) and full combustion (FC) followed by scintillation detection have been applied for determining the tritium activity concentrations in JET divertor tiles. These methods have two main differences – the range of detection and the spatial resolution – and are thus complementary. However, these differences can also complicate the comparison of the two techniques for typical JET divertor samples. Therefore a cross comparison exercise for tritium measurements was performed between the two methods using specially produced identical standard samples. The cross comparison measurements were performed firstly on the identical standard samples and were extended then to selected samples from the JET divertor tile (laser treated and non-treated samples from tile 14ING3B exposed in 2001–2004). The results obtained by AMS and FC agreed within an error limit of 10% for the standard samples and of 18% for the JET divertor samples. The results of this comparison study provided also useful information concerning the efficiency of tritium removal from the plasma facing surface of the divertor tiles by means of laser ablation.

Transient Thermal and Structural Mechanics Investigation of the New Solid Tungsten Divertor Tile for Special Purposes at ASDEX Upgrade

A new solid tungsten divertor for the fusion experiment axial symmetric divertor experiment upgrade is under construction at present. For special purposes of the plasma diagnostic in the divertor region, special formed solid tungsten divertor tiles are required. A so-called Langmuir probe is used to determine the ion temperature, ion density, and ion potential of the plasma. With the aim to place the probe on the right position, some of the divertor tiles (nine at the device circumference) have been adequately adapted. This paper discusses the main results of the numerical analysis of the thermomechanical behavior under heat load. Initially, the elastic-plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. The influence of a possible material degradation due to thermal cracking was studied. Additionally, the knowledge gained by the numerical analysis was used for the shape optimization of the divertor tile. The first results from the numerical life cycle analysis of the tungsten tiles are reported. Finally, based on the knowledge gained by the numerical analysis, in the light of problem complexity, it is recommended to perform some additional thermal tests. These tests should be performed with the aim to increase the reliability of the special-shaped tungsten tile during operation.

Collinear double-pulse laser-induced breakdown spectroscopy as an in-situ diagnostic tool for wall composition in fusion devices

Our recent investigations have indicated the potential of double-pulse laser-induced breakdown spectroscopy (DP-LIBS) for analysis of the deposited impurity on divertor tiles in the experimental advanced superconducting tokamak (EAST). The collinear DP-LIBS configuration was suggested with the aim of overcoming the sensitivity shortcomings of the conventional single pulse laser-induced breakdown spectroscopy (SP-LIBS) technique in vacuum conditions. A systematic study of plasma emission signal dependence on the inter-pulse delay at 3.5×10−3Pa was performed, and the results were compared with the ones obtained with a single laser pulse of energy corresponding to the sum of the two pulses. For a molybdenum tile, it was found that the atomic spectral lines of Mo were enhanced by a factor of 6.5 when an inter-delay time of 1.5μs was installed. For an exposed divertor tile, significant increases in the emission line intensities of various minor elements (such as Mo, Si, Fe, Cr, Ti, Ni and Ca) were observed in DP-LIBS, while no spectra signal was achieved in SP-LIBS. This collinear DP-LIBS technique would help us to develop a more promising system to monitor the fuel retention and impurity deposition on plasma facing components of EAST.