Adhesive force distributions for tungsten dust deposited on bulk tungsten and beryllium-coated tungsten surfaces

Comprehensive measurements of the adhesive force for tungsten dust adhered to tungsten surfaces have been performed with the electrostatic detachment method. Monodisperse spherical dust has been deposited with gas dynamics techniques or with gravity mimicking adhesion as it naturally occurs in tokamaks. The adhesive force is confirmed to follow the log-normal distribution and empirical correlations are proposed for the size-dependence of its mean and standard deviation. Systematic differences are observed between the two deposition methods and attributed to plastic deformation during sticking impacts. The presence of thin beryllium coatings on tungsten surfaces is demonstrated to barely affect adhesion

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

Proceedings of the 22nd International Conference on Plasma Surface Interactions 2016, 22nd PSIAs a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.Peer reviewe

Raman microscopy to characterize plasma-wall interaction materials: from carbon era to metallic walls

Plasma-wall interaction in magnetic fusion devices is responsible for wall changes and plasma pollution with major safety issues. It is investigated both in situ and ex situ, especially by realizing large scale dedicated post-mortem campaigns. Selected parts of the walls are extracted and characterized by several techniques. It is important to extract hydrogen isotopes, oxygen or other element content. This is classically done by ion beam analysis and thermal desorption spectroscopy. Raman microscopy is an alternative and complementary technique. The aim of this work is to demonstrate that Raman microscopy is a very sensitive tool. Moreover, if coupled to other techniques and tested on well-controlled reference samples, Raman microscopy can be used efficiently for characterization of wall samples. Present work reviews long experience gained on carbon-based materials demonstrating how Raman microscopy can be related to structural disorder and hydrogen retention, as it is a direct probe of chemical bonds and atomic structure. In particular, we highlight the fact that Raman microscopy can be used to estimate the hydrogen content and bonds to other elements as well as how it evolves under heating. We also present state-of-the-art Raman analyses of beryllium- and tungsten-based materials, and finally, we draw some perspectives regarding boron-based deposits.</p

RF Discharge Mirror Cleaning for ITER Optical Diagnostics Using 60 MHz Very High Frequency

For the fusion reactor ITER, a mandatory monitor of the fusion device and plasma will be performed with optical diagnostic systems. For the metallic first mirrors, the recovery of the reflectivity losses due to dust deposition is proposed to be carried out for 14 different optical diagnostic systems by the plasma cleaning technique. In this work, we studied the influence of the electrode area on the electrode potential as a function of the applied power with a 60 MHz radio very high frequency source. Unshielded copper disks with different diameters were constructed to study the impact of the electrode area in the range of 90 cm2 to 1200 cm2, which corresponds to an Edge Thomson Scattering area ratio of 0.15 to 2. It was observed that the absolute value of the resulting bias decreased from 280 V to 15 V with the increase of the area for a given RF power. Moreover, the power consumption was reduced by 43 langid = english, keywords = End-of-Cleaning indicator,First mirror,ITER,Plasma cleanin

Analysis of retained deuterium on Be-based films: Ion implantation vs. in-situ loading

Pure Be, Be-O and Be-O-C thin coatings were deposited using high-power impulse magnetron sputtering (HiPIMS) with and without incorporation of deuterium. The coatings produced without deuterium were implanted afterwards with 15 keV 2H+ ion beams with a fluence limited to 2 × 1017 ion/cm2 in order to mitigate the damage imposed by ion irradiation and prevent a fast gas release. The as-deposited and as-implanted coatings were analysed by IBA techniques, namely by elastic and Rutherford backscattering spectrometries (EBS and RBS, respectively), nuclear reaction analysis (NRA) and by time-of-flight elastic recoil detection analysis (ToF-ERDA). Despite distinct deuterium depth profiles in the implanted samples, the results show that for the present ion implantation and deposition parameters, similar retained amounts are revealed in the films loaded by ion implantation or during the HiPIMS deposition, assuring ion implantation as a competitive and reliable method for fuel incorporation in thin Be-based films for retention studies in controlled conditions

Quartz micro-balance and in situ XPS study of the adsorption and decomposition of ammonia on gold, tungsten, boron, beryllium and stainless steel surfaces

Gas seeding is often used in tokamaks to reduce the power load onto the divertor target plates. Nitrogen is the preferred seeding species because of its favourable radiative properties as well as its apparent beneficial effect on plasma confinement. However, nitrogen molecules are chemically reactive with hydrogen and its isotopes to form stable ammonia compounds. Since ammonia is a polar molecule, sticking on metal surfaces can be expected, increasing as a consequence the tritium retention which could pose a serious risk for ITER operation and maintenance. It is, therefore, important to understand the adsorption mechanism of ammonia on surfaces, investigate when the surface saturation occurs and whether ammonia adsorbs as a molecule or undergoes a dissociation on the surface. In this contribution, ammonia sticking on different fusion-relevant materials is presented. The results show a pressure-dependent ammonia sticking on tungsten, boron and stainless steel followed by a partial desorption from these surfaces while on gold and beryllium, ammonia molecules weakly adsorb and completely desorb. A detailed explanation of the two interaction mechanisms is addressed. Furthermore, the time dependence of ammonia desorption as well as the chemical state of non-desorbed residuals were investigated with x-ray photoelectron spectroscopy. Tungsten, boron and stainless steel surfaces showed a continuous dissociation process from NH3 to NH2, NH, N and surface nitrides

Plasma-wall interaction studies within the EUROfusion consortium: Progress on plasma-facing components development and qualification

This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.European Commission; Consortium for Ocean Leadership 633053; Institute of Solid State Physics, University of Latvia as the Center of Excellence has received funding from the European Union’s Horizon 2020 Framework Programme H2020-WIDESPREAD-01-2016-2017-TeamingPhase2 under grant agreement No. 739508, project CAMART

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate