Thermal hydraulic analysis of IFMIF lithium target with incident deuteron beam

Thermohydraulische Analyse des IFMIF Lithiumtargets mit einfallendem Deuteronenstrahl In der vorgeschlagenen "International Fusion Materials Irradiation Facility (IFMIF)" soll eine intensive hochenergetische Neutronenquelle über die Stripping-Reaktion mit 30 - 40 MeV Deuteronen in einem Lithiumtarget realisiert werden. Das IFMIF-Target, das einen Wärmeeintrag bis zu 10 MW abführen muß, soll aus einem Flüssig-Lithium-Strahl mit einer freien Oberfläche (in Richtung Deuteronenstrahl) und einer Rückwand (in Richtung Testzellen) bestehen. In dieser Auslegung sind die thermische und hydraulische Stabilität des Lithiumstrahls sowie die Verdampfung von Lithium an der freien Oberfläche von besonderer Bedeutung. Hauptgegenstand der vorliegenden Arbeit war die thermohydraulische Analyse des Lithiumtargets mit einfallendem Deuteronenstrahl. Turbulente Strömung und Wärmeübergang in der Targetdüse und im Lithiumstrahl bei einfallendem Deuteronenstrahl wurden simuliert mit Hilfe des Finite-Elemente-Codes FIDAP. Das Wärmeeintragsprofil infolge Deuteronenabbremsung wurde mit Modellen für die Stopping Power durch Elektronen- und nukleare Wechselwirkung unter Berücksichtigung der Temperaturverteilung im Target bestimmt. Nach unserer Kenntnis ist dies eine der wenigen Studien, die eine konsistente Simulation der turbulenten Strömung sowohl in der Düse als auch im Strahl enthält und auf alle wichtigen Targetkonfigurationen anwendbar ist. Obwohl ein anderer Simulationscode für den Lithiumstrahl und ein anderer Datensatz für die Lithiumeigenschaften benutzt wurde, sind die meisten unserer Ergebnisse für das IFMIF-Referenztarget konsistent mit denen der IFMIF-Partner. Die Verdampfung von der Oberfläche erwies sich als relativ gering für die untersuchten Parameter des Lithium- und Deuteronenstrahls. Die freie Lithiumoberfläche am unteren Ende des Deuteronenstrahls wurde als der kritischste Bereich für eventuelle Siedevorgänge identifiziert. Für relativ niedrige Druckwerte in der Vakuumkammer (kleiner als p = 10-3 Pa) ist festzustellen, daß der Abstand zum Siedepunkt an der freien Oberfläche (z. B. Tb 7°C für p = 10-4 Pa) der begrenzende Faktor für die Targetauslegung darstellt. Eventuelles Sieden innerhalb des Targets ist stark vom Targetkonzept abhängig. Beim Konzept mit gekrümmter Rückwand kommt es durch die Einwirkung der Zentrifugalkraft innerhalb des Lithiumstrahls zu einem quasilinearen Druckaufbau vom Vakuumkammerdruck bis zu p = 1,3 x 104 Pa nahe der Rückwand, wodurch Siedevorgänge innerhalb des Strahles stark unterdrückt werden. Bei den beiden anderen Konzepten, nämlich dem Target mit gerader Rückwand (das einen begrenzten Druckaufbau erlaubt) und dem Freistrahltarget ohne Rückwand (dadurch auch kein innerer Druckaufbau), können Siedevorgänge nur vermieden werden durch eine signifikante Steigerung der Strahlgeschwindigkeit und damit des gesamten Lithiuminventars. Der Einfluß der verdampften Lithiummenge, d. h. ihre Ablagerung und ihre Wechselwirkung mit dem Deuteronenstrahl, bedürfen weiterer Untersuchung. Dazu schlagen wir die Anwendung eines Monte-Carlo-Codes für direkte Simulation vor, um den Masse- und Wärmetransport im Reaktionsraum des Targets und in den Bereichen der Target-Beschleuniger-Wechselwirkung zu modellieren. Schließlich kommen wir, unterstützt durch die Ergebnisse anderer Untersuchungen, zu der Schlußfolgerung, daß unsere Resultate die physikalische Realisierbarkeit des IFMIF-Targets in der Ausführung mit gekrümmter Rückwand bestätigen

The DEMO magnet system – Status and future challenges

We present the pre-concept design of the European DEMO Magnet System, which has successfully passed the DEMO plant-level gate review in 2020. The main design input parameters originate from the so-called DEMO 2018 baseline, which was produced using the PROCESS systems code. It defines a major and minor radius of 9.1 m and 2.9 m, respectively, an on-axis magnetic field of 5.3 T resulting in a peak field on the toroidal field (TF) conductor of 12.0 T. Four variants, all based on low-temperature superconductors (LTS), have been designed for the 16 TF coils. Two of these concepts were selected to be further pursued during the Concept Design Phase (CDP): the first having many similarities to the ITER TF coil concept and the second being the most innovative one, based on react-and-wind (RW) Nb3Sn technology and winding the coils in layers. Two variants for the five Central Solenoid (CS) modules have been investigated: an LTS-only concept resembling to the ITER CS and a hybrid configuration, in which the innermost layers are made of high-temperature superconductors (HTS), which allows either to increase the magnetic flux or to reduce the outer radius of the CS coil. Issues related to fatigue lifetime which emerged in mechanical analyses will be addressed further in the CDP. Both variants proposed for the six poloidal field coils present a lower level of risk for future development. All magnet and conductor design studies included thermal-hydraulic and mechanical analyses, and were accompanied by experimental tests on both LTS and HTS prototype samples (i.e. DC and AC measurements, stability tests, quench evolution etc.). In addition, magnet structures and auxiliary systems, e.g. cryogenics and feeders, were designed at pre-concept level. Important lessons learnt during this first phase of the project were fed into the planning of the CDP. Key aspects to be addressed concern the demonstration and validation of critical technologies (e.g. industrial manufacturing of RW Nb3Sn and HTS long conductors, insulation of penetrations and joints), as well as the detailed design of the overall Magnet System and mechanical structures

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

On the mechanisms governing gas penetration into a tokamak plasma during a massive gas injection

A new 1D radial fluid code, IMAGINE, is used to simulate the penetration of gas into a tokamak plasma during a massive gas injection (MGI). The main result is that the gas is in general strongly braked as it reaches the plasma, due to mechanisms related to charge exchange and (to a smaller extent) recombination. As a result, only a fraction of the gas penetrates into the plasma. Also, a shock wave is created in the gas which propagates away from the plasma, braking and compressing the incoming gas. Simulation results are quantitatively consistent, at least in terms of orders of magnitude, with experimental data for a D 2 MGI into a JET Ohmic plasma. Simulations of MGI into the background plasma surrounding a runaway electron beam show that if the background electron density is too high, the gas may not penetrate, suggesting a possible explanation for the recent results of Reux et al in JET (2015 Nucl. Fusion 55 093013)

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

The effect of beryllium oxide on retention in JET ITER-like wall tiles

Preliminary results investigating the microstructure, bonding and effect of beryllium oxide formation on retention in the JET ITER-like wall beryllium tiles, are presented. The tiles have been investigated by several techniques: Scanning Electron Microscopy (SEM) equipped with Energy Dispersive X-ray (EDX), Transmission Electron microscopy (TEM) equipped with EDX and Electron Energy Loss Spectroscopy (EELS), Raman Spectroscopy and Thermal Desorption Spectroscopy (TDS). This paper focuses on results from melted materials of the dump plate tiles in JET. From our results and the literature, it is concluded, beryllium can form micron deep oxide islands contrary to the nanometric oxides predicted under vacuum conditions. The deepest oxides analyzed were up to 2-micron thicknesses. The beryllium Deuteroxide (BeOxDy) bond was found with Raman Spectroscopy. Application of EELS confirmed the oxide presence and stoichiometry. Literature suggests these oxides form at temperatures greater than 700 °C where self-diffusion of beryllium ions through the surface oxide layer can occur. Further oxidation is made possible between oxygen plasma impurities and the beryllium ions now present at the wall surface. Under Ultra High Vacuum (UHV) nanometric Beryllium oxide layers are formed and passivate at room temperature. After continual cyclic heating (to the point of melt formation) in the presence of oxygen impurities from the plasma, oxide growth to the levels seen experimentally (approximately two microns) is proposed. This retention mechanism is not considered to contribute dramatically to overall retention in JET, due to low levels of melt formation. However, this mechanism, thought the result of operation environment and melt formation, could be of wider concern to ITER, dependent on wall temperatures