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A micro-structuring of the tungsten plasma-facing surface can strongly reduce near surface thermal stresses induced by ELM heat fluxes. This approach has been confirmed by numerical simulations with the help of ANSYS software. For experimental tests, two 10 × 10 mm2 samples of micro-structured tungsten were manufactured. These consisted of 2000 and 5000 vertically packed tungsten fibres with dimensions of Ø240 µm × 2.4 mm and Ø150 µm × 2.4 mm, respectively. The 1.2 mm bottom parts of the fibres are embedded in a copper matrix. The top parts of the fibres have gaps about of 10 µm so they are not touching each others. The top of all tungsten fibres was electro-polished. A Nd:YAG laser with a pulse duration 1 ms and a pulse repetition frequency of 25 Hz was used to simulate up to 105 ELM-like heat pulses. No damage on either of the micro-structured tungsten samples were observed. Neon plasma erosion rate and fuel retention of the micro-structured tungsten samples were almost identical to bulk tungsten samples. Keywords: Micro-structured tungsten, PFC, PFM, High heat load, Thermal cycling, Retention, Erosion, Emissivity

Tungsten is an outstanding material and due to its properties like highest melting point and tensile strength of all natural metals and its high thermal conductivity it is a prime candidate for being used in very harsh environments and for challenging applications like X-ray tubes or as plasma facing material (PFM) in fusion reactors. Unfortunately, high brittle to ductile transition temperature and hardness represent a great challenge for classic manufacturing processes. Additive manufacturing (AM) of tungsten could overcome these limitations and resulting design restrictions. However, AM of tungsten also poses challenges in particular related to the production of material of high density and mechanical stability. Using a selective electron beam melting and a base temperature of 1000 °C of the powder, we were able to produce tungsten with a theoretical density of 99 % without the need of any post-treatment like a second melting step or a redensification by e.g. hot isostatic pressing (HIP). The surface morphology, microstructure, hardness, thermal conductivity and stability against severe transient heat loads were investigated with respect to the relevant building parameters and compared with recrystallized standard W. Besides simple test geometries also more sophisticated ones like monoblocks were successfully realized illustrating the potential of AM for fusion.

Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide an important safety advantage compare to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen preventing the formation of volatile and radioactive WO3. In this work, Zr is added to self-passivating W-10Cr-0.5Y alloy, manufactured by mechanical alloying and HIP, in view of improving its mechanical strength and thus, its thermal shock resistance. The as-HIPed W-10Cr-0.5Y-0.5Zr exhibits a nanocrystalline microstructure with the presence of an extremely fine nanoparticle dispersion. After heat treatment at 1555 °C for 1.5 h, the grain size growths from less than 100 nm to 620 nm and nanoparticles are present both at the grain boundaries and inside the grains. Oxidation tests at 1000 °C revealed that the alloy with Zr exhibits also a strong oxidation reduction compared to pure W. The long-term oxidation rate is similar to that of the alloy without Zr. Under thermal shock loading simulating 1000 ELM-like pulses at the divertor, the heat treated Zr-containing alloy did not present any damage.

The plasma facing first wall in ITER will be armored with beryllium. During operation, the armor has to sustain direct plasma contact during the start-up and ramp-down of the plasma. On top, transient thermal loads originating from a variety of plasma instabilities or mitigation systems are impacting the 8–10 mm thick beryllium tiles. In this work, possible armor thickness losses caused by the expected transient heat loads are reviewed. Applying conservative assumptions, vertical displacement events can cause locally a melt layer with a thickness of up to 3 mm. However, cracks after solidification/cool down are confined to the melt layer and the connection between melt layer and bulk remains strong. Radiative cooling mechanisms can be applied to significantly decrease the melt and evaporation layer thickness. To mitigate the critical damage potential of plasma disruptions, massive gas injections or shattered pellet injections can be deployed to transform the stored plasma energy into radiation, which implies a much more homogeneous distribution of energy to the plasma facing components. For a full power plasma discharge in ITER, these radiative loads can cause temperatures exceeding the melting temperature of beryllium. Experiments have demonstrated that a thickness of 340 µm at the entire first wall armor can be affected by these mechanisms over the lifetime of ITER. Edge localized modes with expected characteristics obtained by fluid model simulations caused fatigue cracks with a depth of up to 350 µm in experimental simulations. The critical heat flux factor FHF above which inflicted damage accumulates with each subsequent pulse has been determined to be in the range of FHF ≈ 9–12 MW m−2 s0.5. The damage from thermal loads below this threshold saturates between 104 and 106 pulses. Neutron irradiation has a deteriorating effect on the thermomechanical properties of beryllium, which strongly influence its resistance against thermally induced damages. The rather low neutron fluence over the lifetime of ITER is expected to reduce the material strength and thermal conductivity by a few tens of percents. If the thickness losses are affected to a similar extent, a sufficient margin of armor thickness will remain. Overall, the damage imposed by radiative loads from massive gas injections or shattered pellet injections is expected to be the dominant force influencing the condition of the first wall armor, at least if all disruptions can be successfully mitigated and the number of vertical displacement events can be constrained to a few occurrences over the service time of ITER. Keywords: ITER, First wall, Plasma facing materials, Beryllium, Erosion, Transient thermal load, ELMs, Massive gas injections, Shattered pellet injections

The experimental fusion reactor ITER will apply beryllium as first wall armor material. In present fusion experiments, e.g. ASDEX Upgrade, it has been detected that up to 25% of the plasma energy loss is deposited in non divertor regions during edge localized mode (ELM) events. Therefore, the impact of transient heating events on beryllium needs to be investigated to reliably predict the performance of the beryllium armor tiles under ITER operational conditions. In the present experiments, the electron beam facility JUDITH 2 was used to exert transient heat pulses with power densities of 0.14–1.0GWm−2, as they can be expected for mitigated Type 1 ELMs in ITER, pulse durations in the range of 0.08–1.0ms, and a number of pulses in the range of 103–107 on S-65 beryllium specimens that were brazed on an actively cooled copper structure. Thereby, a strong drop of the melting threshold was discovered from a heat flux factor FHF = 22–25MWm−2s0.5 for 102 pulses to FHF

In the full-tungsten divertor qualification program at ITER Organization, macro-cracks, so called self-castellation were found in a fraction of tungsten monoblocks during cyclic high heat flux loading at 20MW/m2. The number of monoblocks with macro-cracks varied with the tungsten products used as armour material. In order to understand correlation between the macro-crack appearance and W properties, an activity to characterize W monoblock materials was launched at the IO. The outcome highlighted that the higher the recrystallization resistance, the lower the number of cracks detected during high heat flux tests. Thermo-mechanical finite element modelling demonstrated that the maximum surface temperature ranges from 1800 °C to 2200 °C and in this range recrystallization of tungsten occurred. Furthermore, it indicated that loss of strength due to recrystallization is responsible for the development of macro-cracks in the tungsten monoblock. Keywords: Fracture, Tungsten, ITER, Divertor, High heat flux, Recrystallization

In the current study, a solid state disc laser with a wavelength of 1030 nm and maximum power of 5.3 kW was used to melt the surface of pure tungsten samples (manufactured according to ITER specifications by Plansee SE). Several combinations of laser power and traverse velocity were tested, with the aim of eliminating any pre-existing cracks and forming a smooth and contiguous resolidified surface. Some of the samples were previously damaged by the electron beam simulation of 100 THLs of 0.38 GW/m² intensity (Δt = 1 ms) on a 4 × 4 mm² area in the JUDITH 1 facility. These conditions were chosen because the resulting damage (crack network) and the crack depth (∼200–300 µm) are known from previous identical material tests with subsequent cross sectioning. After laser melting, the samples were analyzed by SEM, laser profilometry and metallographic cross sectioning. A closed surface without cracks, an increased grain size and pronounced grain boundaries in the resolidified area were found. Profilometry proved that the surface height variations are within ±25 µm from the original surface height, meaning a very smooth surface was achieved. These results successfully demonstrate the possibility of repairing a cracked tungsten surface by laser surface re-melting. This “laser repair” could be used to extend the lifetime of future plasma facing components.

One of the numerous challenges of the demonstration power plant DEMO is the selection of appropriate plasma facing materials (PFMs) and this task is ultimately important to the success for DEMO. Low-activation stainless steel (e.g. EUROFER, P92), which is already intended as structural material, could also become a possible plasma facing material, e.g. for the first wall (FW). Therefore, the ferritic martensitic steel P92 was investigated under DEMO relevant loading conditions. An area of the sample surfaces was firstly molten by transient events with varying power densities (A = 245 MW/m 2 , B = 708 MW/m 2 ) and afterwards simultaneously and sequentially exposed to thermal and particle loads. Surface modifications and pronounced microstructure changes were investigated dependent on the pre-exposure, loading sequence and power density. More precisely, it turned out that there was no connection between the loading sequence and the surface modifications for the preloaded A-samples contrary to preloaded B-samples. The preloaded B-samples exhibited surface roughening, melting and the formation of holes dependent on the loading sequence and power density.

Massive gas injections (MGIs) will be used in ITER to mitigate the strong damaging effect of full performance plasma disruptions on the plasma facing components. The MGI method transforms the stored plasma energy to radiation that is spread across the vacuum vessel with poloidal and toroidal asymmetries. This work investigated the impact of MGI like heat loading on the first wall armor material beryllium. ITER-relevant power densities of 90-260 MW m−2 in combination with pulse durations of 5-10 ms were exerted onto the S-65 grade beryllium specimens in the electron beam facility JUDITH 1. All tested loading conditions led to noticeable surface morphology changes and in the expected worst case scenario, a crater with thermally induced cracks with a depth of up to ∼340 µm formed in the loaded area. The level of destruction in the loaded area was strongly dependent on the pulse number but also on the formation of beryllium oxide. The cyclic melting of beryllium could lead to an armor thinning mechanism under the presence of melt motion driving forces such as surface tension, magnetic forces, and plasma pressure.

Experiments were performed in the linear plasma device PSI-2 in order to investigate the synergistic effects of combined steady-state He-plasma and thermal shock exposure. Tungsten produced according to the ITER material specifications by Plansee SE, Austria, was loaded sequentially and simultaneously by steady-state He plasma and transient thermal loads induced by a laser beam. All tungsten samples were exposed to helium plasma for 40 min at a base temperature of ca. 850 °C and a flux of ca. 2.8 × 1022 m−2s−1. Before, during and after the plasma exposure 1000 thermal shock pulses with a pulse duration of 1 ms and a power density 0.76 GW/m² were applied on the samples. The thermal shock exposure before and after plasma exposure was done at room temperature in order to investigate helium induced surface effects also within cracks. The obtained results show that the combination of He plasma with transient thermal shock events results in a severe modification such as reduced height or agglomeration of the sub-surface He-bubbles and of the created nanostructures, i.e. W-fuzz.

Five samples of recrystallized pure tungsten were exposed to transient heat loads using the electron beam of the JUDITH 1 and JUDITH 2 installations of Forschungszentrum Julich. The heat flux and base temperature were the same for all samples; only the number of pulses and exposure device differed. Transmission electron microscopy was applied to determine the first defects that are introduced during exposure and to compare the effects of the two machines. With increasing number of pulses, first dislocations are formed near the grain boundaries, and then line dislocations and clusters of dislocations appear within the grains. Upon prolonged exposure, the dislocations migrate and cluster in dislocation pile-ups. Comparing exposure in JUDITH 1 to JUDITH 2, the amount of defects is much higher in the samples exposed in JUDITH 1.

Structural damage due to high flux particle irradiation can result in significant changes to the thermal strength of the plasma facing component surface (PFC) during off-normal events in a tokamak. Low-energy He+ ion irradiation of tungsten (W), which is currently the leading candidate material for future PFCs, can result in the development of a fiber form nanostructure, known as “fuzz”. In the current study, mirror-finished W foils were exposed to 100eV He+ ion irradiation at a fluence of 2.6 ×1024ionsm−2 and a temperature of 1200K. Then, samples were exposed to two different types of pulsed heat loading meant to replicate type-I edge-localized mode (ELM) heating at varying energy densities and base temperatures. Millisecond (ms) laser exposure done at 1200K revealed a reduction in fuzz density with increasing energy density due to the conglomeration and local melting of W fibers. At higher energy densities (∼ 1.5MJm−2), RT exposures resulted in surface cracking, while 1200K exposures resulted in surface roughening, demonstrating the role of base temperature on the crack formation in W. Electron beam heating presented similar trends in surface morphology evolution; a higher penetration depth led to reduced melt motion and plasticity. In situ mass loss measurements obtained via a quartz crystal microbalance (QCM) found an exponential increase in particle emission for RT exposures, while the prevalence of melting from 1200K exposures yielded no observable trend. Keywords: Plasma facing materials, Fuzz formation, Tungsten, ELM, ITER, Transient heat loading

The investigation of plasma facing materials (PFM) subjected to a large number (≥10,000) of thermal shocks is of interest to determine long term morphological changes which might influence component lifetime in and plasma performance of a fusion reactor. The electron beam facility JUDITH 2 was used to simulate these conditions experimentally. In this study eight different tungsten grades produced by powder injection molding (PIM) were investigated: Two pure tungsten grades, one with 2 wt% Y2O3, three with 1, 2 and 3 wt% TiC, and two with 0.5 and 1 wt% TaC. Samples of 10 × 10 × 4 mm³ were brazed to a copper cooling structure and subjected to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW/m² (FHF = 12 MWs½/m2) at a base temperature of Tbase = 700 °C.The PIM grades showed damages in general comparable with a sintered and forged pure tungsten reference grade (>99.97 wt% W) that complies with the ITER specifications. One exception was the 2 wt% TiC doped material which failed early during the experiment by delamination of a large part of the surface. The Y2O3 doped material showed a comparatively good performance with respect to crack width (

Since the ITER divertor design includes tungsten monoblocks in the vertical target where heat loads are maximal, the design to protect leading edges as well as technology R&D for high performance armor-heat sink joint were necessary to be implemented. In the R&D, the availability of the technology was demonstrated by high heat flux test of tungsten monoblock components. Not systematically but frequently macro-cracks appeared at the middle of monoblocks after 20 MW/m2 loading. The initiation of such macro-cracks was considered to be due to cyclic exposure to high temperature, ∼2000 °C, where creep, recrystallization and low cycle fatigue were concerned. To understand correlation between the macro-crack appearance and mechanical properties and possible update of acceptance criteria in the material specification, an activity to characterize the tungsten monoblocks was launched.

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER. For the realization of fusion energy especially materials questions pose a significant challenge already today. Advanced materials solution are under discussion in order to allow operation under reactor conditions [1] and are already under development used in the next step devices. Apart from issues related to material properties such as strength, ductility, resistance against melting and cracking one of the major issues to be tackled is the interaction with the fusion plasma. Advanced tungsten (W) materials as discussed below do not necessarily add additional lifetime issues, they will, however, add concerns related to erosion or surface morphology changes due to preferential sputtering. Retention of fuel and exhaust species are one of the main concerns. Retention of hydrogen will be one of the major issues to be solved in advanced materials as especially composites and alloys will introduce new hydrogen interactions mechanisms. Initial calculations show these mechanisms. Especially for Helium as the main impurity species material issues arise related to surfaces modification and embrittlement. Solutions are proposed to mitigate effects on material properties and introduce new release mechanisms.