1,564 research outputs found

    Melt infiltrated Tungsten-Copper composites as advanced heat sink materials for plasma facing components of future nuclear fusion devices

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    The exhaust of power and particles is regarded as a major challenge in view of the design of a magnetic confinement nuclear fusion demonstration power plant (DEMO). In such a reactor, highly loaded plasma facing components (PFCs), like the divertor vertical targets, have to withstand both severe high heat ux loads and considerable neutron irradiation. Existing divertor target designs make use of monolithic tungsten (W) and copper (Cu) material grades that are combined in a PFC. Such an approach, however, bears engineering difficulties as W and Cu are materials with inherently different thermomechanical properties and their optimum operating temperature windows do not overlap. Against this background, W-Cu composite materials are promising candidates regarding the application to the heat sink of highly loaded PFCs. The present contribution summarises recent results regarding the manufacturing and characterisation progress of such W-Cu composite materials produced by means of liquid Cu melt infiltration of open porous W preforms. On the one hand, this includes composites manufactured by infiltrating powder metallurgically produced W skeletons. On the other hand, W-Cu composites based on textile technologically produced fibrous reinforcement preforms are discussed

    Fast low-temperature irradiation creep driven by athermal defect dynamics

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    The occurrence of high stress concentrations in reactor components is a still intractable phenomenon encountered in fusion reactor design. We observe and quantitatively model a non-linear high-dose radiation mediated microstructure evolution effect that facilitates fast stress relaxation in the most challenging low-temperature limit. In situ observations of a tensioned tungsten wire exposed to a high-energy ion beam show that internal stress of up to 2 GPa relaxes within minutes, with the extent and time-scale of relaxation accurately predicted by a parameter-free multiscale model informed by atomistic simulations. As opposed to conventional notions of radiation creep, the effect arises from the self-organisation of nanoscale crystal defects, athermally coalescing into extended polarized dislocation networks that compensate and alleviate the external stress.Comment: 10 pages, 5 figure

    Summary of the 3rd IAEA technicalmeeting on divertor concepts

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    This report summarizes the contributions presented at the 3rd IAEA technical meeting on divertor concepts, held in Vienna, Austria, 4–7 November 2019. The meeting brought together more than 70 experts from nuclear fusion research sites worldwide to discuss the different aspects that the divertor design and fusion machine operation involve, from ITER divertor developments to innovative technologies for future DEMO divertor. The main topics of the meeting were: divertor and confinement; radiative power exhaust; scrape-off layer (SOL) and divertor physics; steady state operation and transient heat loads; plasma facing components materials and heat exhaust for steady state operation; and divertors for DEMO and future power reactors

    Minimally Invasive Vacuum-Assisted Closure Therapy With Instillation (Mini-VAC-Instill) for Pleural Empyema

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    Enthusiasm for minimally invasive thoracic surgery is increasing. Thoracoscopy plays a significant therapeutic role in the fibrinopurulent stage (stage II) of empyema, in which loculated fluid cannot often be adequately drained by chest tube alone. For some debilitated and septic patients, further procedures such as open-window thoracostomy (OWT) with daily wound care or vacuum-assisted closure (VAC) therapy are necessary. In the present article, we propose a new option of minimally invasive VAC therapy including a topical solution of the empyema without open-window thoracostomy (Mini-VAC-instill). Three patients who underwent surgery using this technique are also presented. The discussion is focused on the advantages and disadvantages of the approach

    Microstructure, oxidation behaviour and thermal shock resistance of selfpassivating W-Cr-Y-Zr alloys

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    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
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