Global SOLPS-ITER and ERO2.0 coupling in a linear device for the study of plasma-wall interaction in helium plasma

Plasma–wall interaction (PWI) is a great challenge in the development of a nuclear fusion power plant. To investigate phenomena like erosion of plasma-facing components, impurity transport and redeposition, one needs reliable numerical tools for the description of both the plasma and the material evolution. The development of such tools is essential to guide the design and interpretation of experiments in present and future fusion devices. This contribution presents the first global simulation of PWI processes in a linear plasma device mimicking the boundary plasma conditions in toroidal ones, including both the description of plasma and impurity transport and of plasma-facing material evolution. This integrated description is obtained by coupling two of the state-of-the-art numerical codes employed to model the plasma boundary and the PWI, namely SOLPS-ITER and ERO2.0. Investigation of helium plasma is also of primary importance due to the role helium will have during ITER pre-fusion power operation, when it is planned to be used as one of the main plasma species, as well as fusion ash in full power operation. The plasma background is simulated by SOLPS-ITER and the set of atomic reactions for helium plasmas is updated, including charge-exchange and radiative heat losses. ERO2.0 is used to assess the surface erosion in the GyM vessel, using different wall materials (e.g. carbon, iron or tungsten) and applying different biasing voltage. Eroded particles are followed within the plasma to assess their redeposition location. The ionization probability of the different materials in the GyM plasma is inferred through the energy distribution of impacting particles and its effects on migration are investigated

Validation of the plasma-wall interaction simulation code ERO2.0 by the analysis of tungsten migration in the open divertor region in the Large Helical Device

Tungsten migration in the open divertor region in the Large Helical Device is analyzed for validating the three-dimensional plasma-wall interaction simulation code ERO2.0. The ERO2.0 simulation reproduced the measurement of localized tungsten migration from a tungsten-coated divertor plate installed in the inboard side of the torus. The simulation also explained the measurement of the high tungsten areal density in the private side on a carbon divertor plate, next to the tungsten-coated divertor plate, by the tungsten prompt redeposition in plasma discharges for a low magnetic field strength in a counterclockwise toroidal direction. However, the simulation disagreed with the measurement of low tungsten areal density on the plasma-wetted areas on the carbon divertor plates, which indicated that the actual erosion rate of the redeposited tungsten should be much higher than that used in the ERO2.0 code

Deposition of 13C tracer and impurity elements on the divertor of Wendelstein 7-X

Carbon impurity transport and deposition were investigated in the Wendelstein 7-X stellarator by injecting isotopically labelled methane ((CH4)-C-13) into the edge plasma during the last plasma operations of its Operational Phase (OP) 1.2B experimental campaign. C-13 deposition was measured by secondary ion mass spectrometry (SIMS) on three upper divertor tiles located on the opposite side of the vessel to the(13)CH(4) inlet. The highest C-13 inventories were found as stripe-like patterns on both sides of the different strike lines. These high deposition areas were also analysed for their impurity contents and the depth profiles of the main elements in the layers. Layered deposition of different impurity elements such as Cr, Ni, Mo and B was found to reflect various events such as high metallic impurities during the OP1.2A and three boronizations carried out during OP1.2B.Peer reviewe

Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m(2). Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma-wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.Peer reviewe

First results from recent JET experiments in Hydrogen and Hydrogen-Deuterium plasmas

The hydrogen campaign completed at JET in 2016 has demonstrated isotope ratio control in JET-ILW using gas puffing and pellets for fuelling, Neutral Beam Injection alone or in combination, with D/H spectroscopy as a diagnostic. The plasma properties such as confinement, L-H threshold, density limit depend on the isotope composition. The L-H transition power increases with the hydrogen concentration with a wide plateau in the range 0.2<nH/(nD+nH)<0.8. Energy confinement is significantly lower in hydrogen than in comparable deuterium ELMy H-mode plasmas, suggesting an isotope mass scaling that is stronger than in IPB98(y,2). In L-mode, the isotope dependence of confinement is weaker. The H-mode density limit in hydrogen is up to 35% lower than in heuterium, whilst it is found to be higher in L-mode. The lower ion mass leads to reduced tungsten sputtering in hydrogen plasmas. During the campaign, the nD/(nD+nH) ratio dropped to ~1% in only a few discharges after the last deliberate introduction of deuterium, although it was seen to rise again to ~2% with several seconds of exposure of the divertor tiles to ~10MW of auxiliary heating. Several ICRH scenarios were also tested in hydrogen plasmas

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