Wall Forum 2021

Net erosion, deposition and material transport in the stellarator W7-X were investigated on the Test Divertor Unit (TDU) using special carbon marker coatings during the operational phases OP1.2a in the year 2017 and OP1.2b in 2018, and on the divertor baffles by analysis of regular wall elements after OP1.2b. While OP1.2a was characterised by high concentrations of oxygen and carbon impurities in the plasmas, OP1.2b had much smaller impurity concentrations due to regular boronizations and showed considerably better plasma performances with higher plasma densities. The samples were analysed by quantitative ion beam analysis methods, scanning electron microscopy, laser-induced breakdown spectroscopy (LIBS), and laser-induced ablation-quadrupole mass spectrometry (LIA-QMS). Very high net erosion of carbon was observed at the strike line of all 10 TDUs in both campaigns and is attributed to sputtering and chemical erosion by C and O impurities in OP1.2a. Re-deposition of carbon in remote areas of the TDU was small, but a noticeable deposition of boron on the TDU was observed after OP1.2b. Thicker re-deposited carbon/boron layers with thicknesses of a few microns having a layered structure (carbon and oxygen in OP1.2a; carbon, boron and higher oxygen contents in OP1.2b) were found on divertor baffle tiles after OP1.2b. These tiles contained also higher H-inventories up to 1E22 H-atoms/m2 due to codeposition. The following inner wall showed net deposition of about 100 nm thick boron/carbon layers due to the boronizations. The global carbon balance is currently unclear. However, carbon eroded from the TDU was not redeposited in the divertor region but was transported out of the divertor area. Some carbon was redeposited at the divertor baffles but was also pumped out as CO, CO2, or CH4. This is a profound difference to divertors in tokamaks, where eroded material is typically redeposited in remote divertor areas or in the inner divertor. [mehr]

Most of the challenges in joining and additive manufacturing of metallic structures are related to metallurgical phenomena or inadequate processing routes. Within this talk, the core areas of research at the chair of materials engineering of additive manufacturing will be presented. Targeting a solid understanding of metallurgical phenomena, influenced by various processing variants allows the derivation of structure-process-property relationships, which can then in turn result in tailored materials and processing. In order to getting a comprehensive understanding of materials structure and processing response, most advanced experimental methods are utilised. This includes for example in-situ synchrotron X-ray diffraction or tomography, high temperature microscopy or thermo-physical material simulation. To foster a better understanding of complex relationships, currently, also data mining and implementation of AI-driven approaches are employed. Exemplarily, this will be shown for creep resistant martensitic steels, the combination of welding and forming of steel structures, dissimilar metal welds between aluminium and copper, the diffusion bonding of molybdenum and tungsten and additive manufacturing and post processing of aluminium alloys. [mehr]

During the last years, the use of high-Z plasma facing-materials (PFCs) in fusion devices renewed the interest in arcing. Arcs are considered as source of dust particles and additional, localized erosion mechanism of the PFCs. In this talk arc investigations at IPP of during the last years will be resumed. Present investigations yields that the erosion by arcs depends strongly on the melting temperature of the substrate and higher permeability enhances the erosion by an order of magnitude. Droplets produced by arcs, which were injected into the plasma direction have a much higher penetration probability than atom and ions. A laboratory device to investigate these was developed and first results were presented. [mehr]

Thermomechanical simulation of the new upper divertor tiles in AUG using ANSYS Workbench . Non-linear modelling was used, including contacts and material properties as a function of the temperature. After the completion of the analyses, benchmarking at GLADIS was performed to verify the behaviour. Finally, normal operation at AUG was modelled to assess the performance and resistance of the tile. [mehr]

Experimental studies of tungsten (W) melt dynamics during repeated ELM transients on both ASDEX Upgrade and JET, together with MEMOS-U melt code analysis, have shown conclusively that melt motion is mainly driven by the Lorentz force due to the magnetic field acting on an electric current passing through the melt layer arising to compensate the thermionic electron emission from the hot plasma-facing surface. To further verify the melt transport model a new experiment has been carried out at ASDEX Upgrade in which transient melting was compared on two samples of identical geometry and at identical exposure positions, but with one sample electrically floating and the other connected to vessel potential. Thermionic emission cannot drive a current on the floating sample, but will instead modify the local sheath potential at the plasma-facing surface. As expected, significant differences were found in the post-exposure morphology of the two samples. Melt motion appears to be weaker at the floating sample surface, supporting the assertion of net replacement current as the main driving mechanism. However, the substantially higher power flux to the floating sample, attributed to an increased local sheath heat transmission factor, complicates the picture, preventing straightforward deductions. Simulations are underway with MEMOS-U to disentangle the two mechanisms. [mehr]

Prof. Große will at first present his chair of Non-destructive Testing (NDT) at the TU Munich. He will introduce his team, speak about teaching and available measuring techniques. After giving an overview of the field of research he is working in he will show modules for efficient NDE (nondestructive evaluation) applications. Finally he will talk about selected projects as for example the rehabilitation of a bridge in Munich, NDE of a ME163b of Flugwerft and very detailed about the archeological investigations of the pyramid of King Khufu (Cheops) in Gizeh. [mehr]

Structural components in commercial fusion reactor designs will be subjected to unprecedented heat and neutron fluxes, potentially leading to severe degradation of their performance. The key challenge of materials modelling is therefore the prediction of thermomechanical properties of structural components over their operational lifetime. I will present recent progress on simulating irradiated microstructure in the Materials Modelling Group at the Culham Centre for Fusion Energy (CCFE), and how statistical physics plays a surprising role in the phenomenon of irradiation embrittlement of components. [mehr]

Das Ziel dieser Arbeit war es, Wolfram-Kupfer-Verbundwerkstoffe auf Basis additiv gefertigter Vorformen herzustellen und zu charakterisieren. Dazu wurden 3D-Modelle sowohl für gitter-basierte als auch wabenbasierte Vorformen erstellt, die anschließend am Fraunhofer-Institut für Gießerei-, Composite- und Verarbeitungstechnik hergestellt wurden. In Vorexperimenten wurden optimale Fertigungsparameter ermittelt und damit Material basierend auf einer Gitter- oder einer Wabenstruktur mit einen Wolframvolumenanteil von 15%, 30% und 45% hergestellt. Neben Zug- und Druckproben wurden auch Proben zur Ermittlung der Temperaturleitfähigkeit gefertigt.In den Zug- und Druckversuchen konnte gezeigt werden, dass der auf Waben basierende Verbundwerkstoff nur für Belastungen parallel zu den Waben geeignet ist. In dieser Richtung hat der Werkstoff die höchste Festigkeit und zeigt gleichzeitig ein duktiles Verhalten beim Bruch. Senkrecht zu den Waben ist die Festigkeit deutlich geringer und der Werkstoff verhält sich sehr spröde. Der Gitter-basierende Verbundwerkstoff kann für alle Belastungsrichtungen eingesetzt werden, zeigt insgesamt aber eine geringere Festigkeit als ein Wabenbasierter in paralleler Lastrichtung. Ab einem Wolfram-Volumenanteil von 35% verhalten sich alle Proben unabhängig von ihrer internen Struktur bei Zugbelastung spröde. Ab diesen Wolfram-Volumenanteil sinkt ebenfalls die Temperaturleitfähigkeit sehr deutlich ab, weshalb dieser Wert für die Verwendung als Wärmesenke nicht überschritten werden sollte. [mehr]

Within European material’s programme, the portfolio of baseline materials for DEMO contains the following items: (i) EUROFER(97), a 9Cr Reduced Activation Ferritic Martensitic (RAFM) steel, as structural material for the breeding blanket, (ii) commercially pure tungsten as plasma facing component armor material, and (iii) copper chromium zirconium (CuCrZr) alloy as heat sink material for the divertor coolant interface. This contribution reviews the efforts done towards the assessment of the irradiation effects and operational temperature window performed over the last several years in the frame of the European fusion programme focusing on the armour and heat sink materials.Based on the already available knowledge gained in FP6 and FP7 programmes, the operational conditions for the baseline in-vessel materials are tentatively determined. For each of the baseline material, the lower temperature bound is defined by the embrittlement (fracture without plastic deformation), while the upper temperature bound is determined by softening of the material (reduction of the yield point). Accordingly, the main challenges in the formulated irradiation programmes were linked to: (I) assessment of the ductile-to-brittle transition temperature (DBTT) of baseline tungsten and advanced tungsten alloys; (ii) investigation of baseline tungsten under irradiation at very high temperature, reflecting operational conditions in divertor; (iii) assessment of the mechanical properties of reinforced CuCrZr alloys. The screening exposure to perform early down selection of the materials was also performed. The choice of the advanced materials is driven naturally by the need to extend the operation temperature/fluence window to extend the design space. Although the fusion neutron spectrum implies an important difference in the transmutation reactions compared to fission spectrum, the current R&D programme utilizes available Material Test Reactors (MTRs).Driven by the technological priorities, the irradiation tests campaigns were arranged in two waves. The first one involved baseline materials (EUROFER97: 20 dpa; tungsten: 1 dpa; CuCrZr: 1 dpa) focusing on delivery of the engineering design data and the second one targeted screening irradiation of the advanced materials (advanced/optimized EUOFER97 specification steels: 2.5 dpa, tungsten alloys and composites: 1 dpa; reinforced CuCrZr: 2.5 dpa). Execution of the programmes was realized in Europe (LVR-15 and BR2 reactors) and USA (HFIR reactor).Extraction of the properties of the neutron exposed materials involved massive post irradiation examination (PIE) campaigns. The performance of the advanced materials for divertor (W and CuCrZr) is assessed and presented, which already at this stage allows drawing some important conclusions. [mehr]

In this talk, the development status of self-passivating W-Cr-Y alloys manufactured by powder metallurgy is reviewed. This development has been started to prevent the potential release of volatile, neutron activated tungsten oxides from a pure W first wall (FW) in case of a loss-of-coolant accident (LOCA) with simultaneous air ingress, where temperatures of the in-vessel components exceeding 1000°C are expected. Properties relevant for the case of a LOCA, as the oxidation resistance, are reviewed. Besides, other material properties relevant for operation under normal conditions are presented, as the thermal and mechanical properties including thermal shock resistant in case of ELMS, resistance to high heat fluxes, thermal stability of the microstructure and D retention among others. Finally, methods for joining these alloys to Eurofer by brazing and by diffusion bonding using HIP are shown. [mehr]

The ITER Test Blanket Module (TBM) is tasked with demonstrating efficient breeding and extraction of tritium inside the environment of a nuclear fusion reactor. In order to meet the strict requirements for tritium self-sufficiency necessary for future fusion power plants, and to address safety concerns and minimize radioactive waste, retained T inventory in and T losses out of the TBM must be accurately assessed. In this work, at the request of F4E, T losses due to retention and permeation in the ITER Water-Cooled Lithium Lead (WCLL) TMB are studied with the aid of the TESSIM-X diffusion-trapping code. Due to the presence of neutron-induced traps in the structural materials, T losses in the TBM are now expected to be a factor of 5 - 10 times larger than what was previously thought. [mehr]

The Creation-Relaxation Algorithm (CRA) [1] has attracted a lot of interest recently as it offers a parameter-free method for generating high-dose (>1 dpa) microstructures using atomistic lattice statics with a clear interpretation of the damage dose. We have shown that this method provides quantitative estimations of experimentally measurable properties [2], and having atomistic detail allows us to watch key processes like loop habit plane rotation and coalescence occur. But the method operates in the zero temperature limit, and the lack of true dynamics means the defect microstructures produced are typically too dense. By contrast, massively overlapping Molecular Dynamics (MD) cascades produce closer estimates to experimental quantities [3], but are so much more expensive that they have been limited to the low dose (<0.1 dpa) regime. Here we show how to link the two methods, using MD cascades to relax CRA simulations, and so achieve the high doses needed. We show how to find the void content of an atomistic simulation, and from this demonstrate we can accurately model deuterium retention measured in nuclear reaction analysis experiments [4]. [1] Phys. Rev. Mater. 4:023605 (2020) [2] Phys. Rev. Lett. 125:225503 (2020) [3] J. Nucl. Mater. 528:151843 (2020) [4] arXiv:2106.12938 (accepted Phys. Rev. Mater 2021) [mehr]

Ions are used to simulate, accelerated-irradiation damage in materials. Most studies focus on displacement damage using self-ions or low-energy protons. Energetic protons between 16 – 30 MeV have the ability to induce combined-displacement and transmutation damage, over macroscopic ranges of 300 – 500 µm in tungsten. However the samples are radioactive post irradiation and radiation protection measures must be followed.Additionally, electronic losses from ions are converted into heat, which form steady state heat loads on the sample. For 16 MeV protons with 10 µA current on a 10 mm diameter sample, the heat loads amounts to ~2 MW/m2. This is simultaneously inflicted on the sample alongside irradiation damage.Pilot irradiations on tungsten have been performed to doses of 0.006 dpa and further analysed with scanning electron microscopy and instrumented indentation. The hardness results are compared against self-ion irradiation, 3 MeV proton irradiation and show similar radiation hardening increase. Transmutation estimates using FISPACT-II were compared against gamma spectroscopy results.The irradiation ideology, methodology and post irradiation results will be detailed and explained in the presentation. [mehr]

In fusion reactors tungsten is exposed to high doses of neutron radiation, which causes an embrittlement of the material. The irradiation leads to microstructural damage and causes transmutation. The microstructural damage can be simulated by heavy ion irradiation. In this work, tungsten ion-radiation was used as an easy to control damage source. Fine grained drawn tungsten fibers with a diameter of 16 µm were used as sample material. Due to limited penetration depth of the ions, the fibers needed to be thinned to a diameter of 5 µm. Oriented towards previous investigations, irradiation experiments with several radiation doses were conducted on the thinned fibers. GIRAFFE, a newly developed device for tensile tests was qualified and used for the evaluation of the irradiated samples. After the tensile tests the diameter of the broken fibers were measured as an indication for their ductility. It was found, that radiation doses up to 10 dpa does not show a measurable influence on the ductility of tungsten fibers. [mehr]

Fuel retention in tungsten (W) as a plasma-facing material, especially of the radioactive hydrogen isotope (HI) tritium, presents severe concerns for operation cost and safety of future fusion devices. In tungsten with very low intrinsic H solubility, HI retention is dominated by trapping at irradiation-induced defects. In our previous work, strong lattice distortion was observed in W surfaces after deuterium (D) plasma exposure with kinetic ion energies significantly below the thresholds for production of stable Frenkel pairs, which caused formation of a D-supersaturated surface layer(D-SSL) containing ~10 at.% of retained D. We recently proposed and experimentally verified a physical model for the SSL production by HI plasmas at sub-threshold ion energy based on hydrogen atom-ion synergy effects. However, the connection between the observed defect microstructures and the unexpectedly high concentrations of retained HIs in the SSL has remained unestablished. In the present work, we exposed W samples to HI plasma and characterized them with transmission electron microscopy (TEM) to determine the defect microstructures in the HI-SSL. High quality TEM thin foil specimens were preparedby adopting a back-thinning electropolishing approach. In planar view, contrast images in kinematical two-beam bright-field conditions confirmed the formation of “black-spot” clusters and their raft structures after a 1×1024 m-2 exposure to D ions of215 eV at 300 K. The average defect size and number density measured 4-5 nm and 1022 m-3, respectively. Since the same SSL defect microstructure forms under H plasma with doubled ion energy as for D plasma exposure, both bulk and foil Wsamples were simultaneously exposed to a series of H ion fluences to track the evolution of the defect microstructure in the HSSL. In order to clarify the correlation between the microstructure and the concentration of retained H, hydrogen depthprofiles were acquired with 1H-15N nuclear reaction analysis on the bulk W samples. The analysis of the defect nature and cluster geometry in the HSSL under different H ion fluences is ongoing. The present work is expected to provide an indepth understanding of HI retention in W materials upon injection of energetic projectiles (ions, charge-exchange neutrals, neutrons) in future fusion devices. [mehr]

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