Search

The cracking thresholds were evaluated for tungsten samples with different microstructure in the course of QSPA Kh-50 repetitive plasma loads. No damage has been observed on the exposed surfaces under 0.1 MJ/m2. Nevertheless, cracks were detected in the bulk of irradiated tungsten (with longitudinal grain orientation). Increasing heat load up to 0.2 MJ/m2 caused the damaging of all types of tungsten targets. The observed cracks propagate to the bulk mainly transversely and parallel to the irradiated surface. The effect of the subsequent exposure with LHD divertor plasma on the tungsten samples was analyzed. The obtained results are discussed.

Material issues pose a significant challenge for future fusion reactors like DEMO. When using materials in a fusion environment a highly integrated approach is required. Damage resilience, power exhaust, as well as oxidation resistance during accidental air ingress are driving issues when deciding for new materials. Neutron induced effects e.g. transmutation adding to embrittlement are crucial to material performance. Here advanced materials such as tungsten fibre-reinforced tungsten Wf/W and fibre-reinforced copper Wf/Cu composites could allow the step towards a fusion reactor. Recent developments in the area Wf/W mark a possible path towards a component mock-up early enough for utilisation in DEMO. High heat-flux tests show that having short fibres at the exposed surface leads to their selective erosion and melting. Initial tests in the linear plasma device PSI-2 confirm this behaviour.

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 and tungsten-based alloys are the leading material choices for the divertor plasma facing components (PFCs) in future fusion reactors. Recrystallization may occur when they undergo high heat loads, drastically modifying the predesigned grain structures and the associated desired mechanical properties. However, the influence of recrystallization on the thermal fatigue behavior of tungsten PFCs still remains unclear. In this study, ITER-grade tungsten was simultaneously exposed to a high-flux hydrogen plasma (∼5 × 1024 m−2 s−1) and high-cycle (104–105) transient heat loads in the linear plasma device Magnum-PSI. By correlating the surface temperature distribution, obtained by analyzing temperature-, wavelength-, and surface-dependent emissivity, and the surface modifications of the plasma exposed specimens, the crack initiation heat flux factor threshold was found to be ∼2 MW m−2 s0.5 (equivalently, ∼0.07 MJ m−2 for a 1 ms pulse). Based on electron backscatter diffraction analyses of cross-sections near the crack initiation sites, faster recrystallization kinetics near the surface compared to literature was observed and the surface cracks preferentially initiated at high angle grains boundaries (HAGBs). Upon recrystallization, the yield strength decreases which entails increasing cyclic plastic strains. The HAGBs fraction is increased, which constrains the transfer of plastic strains at grain boundaries. The recrystallization decreases the dislocation density, which promotes heterogeneous deformation. All these mechanisms explain the reduced crack initiation threshold of recrystallized tungsten compared to its as-received counterpart. The results provide new insights into the structural failure mechanisms in tungsten PFCs exposed to extreme fusion plasmas.

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.

The differences in the thermophysical properties between tungsten and steel create thermal stress peaks at its interface when joined directly for the breeding blanket of a fusion reactor. In order to solve this problem, a graded interlayer made of several layers of W/steel-composites is considered to reduce these stress peaks. Plasma spraying under an argon shrouded chamber was employed as a cost efficient manufacturing technique for such composites and field assisted sintering technology/spark plasma sintering was used to diffusion bond them with bulk-W and bulk-steel. Firstly, thermophysical characterizations were performed on these composites. Secondly, two approaches have been investigated to join bulk-W and 75 vol% W composite: direct joining and using a vanadium foil. Only vanadium foil resulted in successful bond formation at all the three bonding temperatures of 800 °C, 900 °C and 1000 °C. Thirdly, investigation of diffusion bonding parameters (temperature and time) for the joining of 25 vol% W, 50 vol% W, 75 vol% W and bulk-steel were studied and optimum process parameter were identified. Finally, this optimized parameter (1000 °C; 30 min) was employed to manufacture a complete 12 mm x 12 mm W-steel joint consisting of this graded interlayer.

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 effect of continued plasma exposure on two divertor target tiles intentionally damaged before installation was studied for ASDEX Upgrade H-mode discharge conditions. On one tile, made of molybdenum alloy (TZM), superficial melt damage was created in the GLADIS high heat flux test facility. To measure the influence of the resulting surface corrugations on erosion, the tile surface was subsequently covered with a 20 nm tungsten marker layer. A second tile, made of tungsten, was pre-damaged by exposure to 2 × 105 consecutive ELM-like heat pulses in the electron beam test facility JUDITH, which resulted in formation of an extended crack network. Both tiles were placed at the outer divertor target of ASDEX Upgrade using the DIM-II divertor manipulator and exposed to a series of 15 identical H-mode discharges. The surface state of both tiles was documented pre- and post-exposure using electron scanning microscopy. SEM analysis of the crack network's microscopic structure did not reveal any additional damage created by plasma exposure but showed deposition of migrated wall materials inside shadowed areas. The erosion pattern of the W-marker layer revealed regions of net erosion as well as of net deposition in the corrugated melt zone. Net erosion up to complete removal of the W marker layer was found at elevated parts of the surface oriented towards the incident plasma flux whereas net deposition was found in corresponding shadowed areas. In contrast, the undamaged surface parts showed a uniform erosion pattern determined by incident plasma ion flux and temperature.

The evaluation of the impact of plasma-facing components damage on subsequent plasma operation is an important issue for ITER. During the first phase of the operation of WEST, a few ITER-like divertor plasma-facing units (PFUs) have been installed on the lower divertor. One PFU was pre-damaged under electron beam gun thermal loading, before its installation in WEST, and the subsequent evolution of the damage was studied after the WEST plasma exposure. This paper presents the procedure followed to get the pre-damaged PFU. It consists of the characterization of the response of tungsten samples representative of WEST PFU under high heat flux (HHF) loading, the selection of damage (namely, small cracks, crack network, crack network, and W melt droplets). Finally, according to the WEST plasma loading conditions, the blocks with damage within the PFU and the position of the pre-damaged PFU on the WEST lower divertor are attributed. The first results obtained after an initial plasma exposure in WEST lead to the assessment, as expected with regard to the heat loading conditions, that no major surface aspect modification was found. This result emphasized the possibility of implementing as pre-damaged small droplets of melted tungsten in a high heat-loaded zone for a future WEST experimental campaign.

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

Microstructure induced by plastic deformation at 600 °C in as-fabricated and recrystallized tungsten grades of ITER specification is studied by means of transmission electron microscopy (TEM). Both grades are produced by AG Plansee and being used by fusion community as research material for plasma facing and high heat flux components. Reference microstructure was investigated by a combination of TEM, scanning electron microscopy and nano-indentation techniques. TEM was applied to investigate the necking region in the tensile samples deformed up to ~ 30% of stain. The deformation-induced microstructure was characterized and compared for the two grades in terms of the dislocation density, heterogeneity, observation of pile-ups and tangles specifically near high angle grain boundaries. The obtained evolution of the dislocation density was compared with the prediction of the previously developed thermo-mechanical model for the plastic deformation of polycrystalline tungsten and a good agreement was found without modification of the original parameter set.

Fusions reactors have to handle numerous specifications before being able to show viable commercial operation, one of which is to find a proper Plasma Facing Material (PFM) which can withstand the high heat loads of several tens of megawatts per square meters combined with the pulse operation of a tokamak and many other problematics (Brezinsek et al 2017 Nucl. Fusion 57 116041). Nowadays, only tungsten is considered as a PFM for high heat flux areas of a tokamak divertor. Tungsten has been selected due to its favorable physical properties, but tungsten has a major drawback: it is brittle under temperatures typically used for water-cooled plasma-facing components (PFC). Under these temperatures the damage threshold due to thermal fatigue induced by ELM is very low, which will dramatically reduce the life-time of the tungsten PFC. The ANSYS simulations and experiments with a millisecond pulsed laser demonstrate a strongly improved ability to withstand thermal fatigue by micro-structuring of the tungsten surface with the help of 150–240 μm diameter tungsten fibres

The ITER divertor will be exposed to extremely high plasma fluences over its lifetime, and it is known that plasma exposure can lead to a variety of particle-induced surface-morphology and microstructure changes in tungsten. However, no data exists at fluences comparable to those expected over extended ITER operations (10 30−31 m−2) and so it is uncertain how these changes will evolve and affect the divertor performance over such long timescales. Six monoblocks were exposed to high flux plasma comparable to partially-detached plasma conditions in the ITER divertor in Magnum-PSI. Different exposures used different plasma species (H, He, D or D + He) and aimed to replicate conditions similar to those during different phases of the ITER staged approach. The highest fluence achieved was 10 30 D m−2, comparable to around one year of ITER Fusion Power Operation. Post-mortem analysis by Nuclear Reaction Analysis revealed very low deuterium retention throughout the blocks, while surface analysis showed no cracking or damage, but did observe helium fuzz growth at low ion energies of 8–18 eV, below typically assumed ion energy requirements for such growth to occur. Metallographic sectioning revealed recrystallization up to 2.2 mm below the surface of monoblocks exposed at peak surface temperatures of up to 1580 °C for different durations up to ~20 h. Finite Element Method analysis coupled to metallographic and Vickers Hardness identification of the boundary of the recrystallized region identified a faster recrystallization process compared to literature expectations, reinforcing that recrystallization dynamics is an important criterion for tungsten grade selection for the ITER divertor. Overall, no major damage or failure was identified, indicating that the design is capable of fulfilling its steady-state performance requirements under high flux, high fluence plasma loading conditions in the ITER divertor.

The development of a divertor target for DEMO is of great importance, being able to sustain the harsh environment that is imposed on this component. To fulfill the loading requirements, different concepts were developed within the EUROfusion WPDIV project. The baseline concept is based on the ITER divertor target W-monoblock design. It is made of tungsten as armour material, CuCrZr as structural material and Cu-OFHC as compliant layer. One of the proposed alternative concepts aims to minimize the stress at interfaces by replacing the thick copper interlayer by W/Cu functionally graded material (FGM). In this study, the FGM interlayer, with a thickness of 500 μm, is composed of stacked elementary layers with the following three compositions: 25 vol.%W + 75 vol.%Cu, 50 vol.%W + 50 vol.%Cu, 75 vol.%W + 25 vol.%Cu. Several FGM interlayers were studied. In total three monoblock type mock-ups were manufactured. This paper describes the steps needed to manufacture mock-ups and characterization of elementary layers (composition, porosity, Young’s modulus). HHF tests applying up to 1000 cycles at 20 MW/m² and sub-sequent post-mortem examinations were performed to qualify the concept performance.

Liquid metal divertors aim to provide a more robust alternative to conventional tungsten divertors. However, they still require a solid substrate to confine the liquid metal. This work proposes a novel design philosophy for liquid metal divertor targets, which allows for a two orders of magnitude reduction of thermal stresses compared to the state-of-the-art monoblock designs. The main principle is based on a 3D-printed tungsten structure, which has low connectedness in the direction perpendicular to the thermal gradient, and as a result also short length scales. This allows for thermal expansion. Voids in the structure are filled with liquid lithium which can conduct heat and reduce the surface temperature via vapor shielding, further suppressing thermal stresses. To demonstrate the effectiveness of this design strategy, an existing liquid metal concept is re-designed, fabricated, and tested on the linear plasma device Magnum-PSI. The thermo-mechanical finite element method analysis of the improved design matches the temperature response during the experiments, and indicates that thermal stresses are two orders of magnitude lower than in the conventional monoblock designs. The relaxation of the strength requirement allows for much larger failure margins and consequently for many new design possibilities.

Tungsten–tantalum (W-Ta) alloys for the application to fusion reactor divertor were developed to overcome the issues of W related to the low temperature brittleness, embrittlement by recrystallization, and embrittlement by neutron irradiation. In the present study, the response to cyclic thermal shock tests as well as the fundamental mechanical properties of W-3%Ta and pure W were investigated, which simulated a transient heat load event in the actual divertor environment. Because it was concerned that the effect of deuterium (D) plasma exposure might be much more pronounced in the Ta-alloyed materials, the thermal shock tests were performed under the background steady state D-plasma exposure. Based on the present study, W-3%Ta might be a competitive alloy system from the viewpoint of thermo-mechanical response in the actual divertor environment as well as the fundamental mechanical properties and resistance to recrystallization.

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.

Knowledge of a material’s thermal stability under extreme synergistic particle and heat loads is crucial for developing high performance reactor materials. In this work, the recrystallization behaviour of tungsten under the influence of hydrogen is investigated by low energy high flux hydrogen plasma exposure for various lengths of time. The microstructural changes following exposure are probed by micro-indentation, electron back-scatter diffraction measurements and the characteristic time for recrystallization is assessed using the Johnson–Mehl–Avrami–Kolmogorov (JMAK) model. A recrystallization activation energy in the range of 425 to 440 kJ.mol-1 is determined, identical to that of oven annealed samples, thereby indicating an insignificant influence of hydrogen plasma on the recrystallization kinetics of tungsten.

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.

Beryllium will be applied as first wall armor material in ITER. The armor has to sustain high steady state and transient power fluxes. For transient events like edge localized modes, these transient power fluxes rise up to 1.0 GW m−2 with a duration of 0.5–0.75 ms in the divertor region and a significant fraction of this power flux is deposited on the first wall as well. In the present work, the reference beryllium grade for the ITER first wall application S-65 was prepared with various surface conditions and subjected to transient power fluxes (thermal shocks) with ITER relevant loading parameters. After 1000 thermal shocks, a crucial destruction of the entire loaded area was observed and linked to the stress accelerated grain boundary oxidation (SAGBO)/dynamic embrittlement (DE) effect. Furthermore, the study revealed that the majority of the thermally induced cracks formed between 1 and 10 pulses and then grew wider and deeper with increasing pulse number. The surface quality did not influence the cracking behavior of beryllium in any detectable way. However, the polished surface demonstrated the highest resistance against the observed crucial destruction mechanism.

The ITER first wall (FW) panels consist of plasma facing Be tiles, the CuCrZr alloy heat sink, and the stainless steel structural material. A copper layer of 1−2 mm is used between the Be tile and the CuCrZr for stress compensation. The results of cyclic high heat flux (HHF) tests employing an electron beam facility indicate that the failure/weak spot usually occurs at the joint corners between the Be tile and the copper layer. 3-D FEM (finite element method) thermo-mechanical analysis utilizing ANSYS software has been conducted to simulate the performance of a FW mock up (MU) under cyclic HHF test. The elastoplastic analysis results indicate that the copper layer is most sensitive to plastic deformation, and eventually the most risky location for failures. With the ANSYS strain-life fatigue module, based on the strain-life relation of copper from the ITER handbook, the fatigue life of the MU was predicted via FEM simulation. The results show that, when loaded with 2 MW/m², the small scale MU and the semi-prototype MU have a predicted life time of 12,220 cycles and 10,834 cycles, respectively. Although the shortest life time is located at the upper surface of copper which is joined to the Be tile, the major part of the copper has a life time of the same scale, which lies in the range of 10000–30000 cycles. When loaded with 2.5 MW/m², the life times of the MUs are 7432 cycles and 7343 cycles, respectively.

International audience; For the development of a divertor for DEMO, the European WPDIV project is underway since 2014. The first phase of the project (2014–2016) aims to provide mock-ups adapted to DEMO operation requirements and the second phase aims to furnish mock-ups, with standardized geometry, which fulfill phase 1 requirements. Within the WPDIV project, several options are under development. One of these aims is to replace the thick copper interlayer, used for ITER divertor components, with a very thin coat (functional gradient material or pure copper) for armor-to-pipe joining. One of the benefits is related to armor temperature which is decreased as the distance of heat conduction path is shortened. Some blocks equipped with thin functional gradient material as interlayer proved, in 2016, to handle high cycling performances without any degradation (no surface change aspect and no decrease of thermal heat exhaust capability). This article gives a brief overview on the recent achievements of the development of thin interlayer concept focusing on the design, mock-up production, inspection, high heat flux (HHF) qualification testing and post-examinations.

International audience; Since tungsten (W) has been selected as material for the ITER divertor target the knowledge of the emissivity in the infrared (IR) wavelength range becomes mandatory. A dedicated setup has been developed to measure the emissivity in the wavelength range 1.7–4.75 μm and the temperature domain 200–850 °C. The paper presents the emissivity measurements on W samples coming from actively cooled bulk W components (ITER-like) and W-coated graphite component of the WEST divertor. W samples with damaged surfaces generated by transient heat load as those observed in fusion machines, including micro-cracks and crack network, are also investigated. The dependence on wavelength, temperature and surface state are shown and discussed.

Due to their high energetic efficiency and therefore ability to cover long distances, underwater gliders have become a common tool for ocean sampling missions. In this work, the application scenario involves multiple underwater gliders which function as payload transfer units to build up an underwater navigation system. Since the gliders are intended to be transported inside of a tubular shuttle system, extensive miniaturization of the vehicles is necessary. Furthermore a low cost design is preferred for inexpensive reproduction. To meet these particular requirements the Micro Glider V2 was developed as a prototype.