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Descripción
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| Tungsten is the main candidate material for the first wall (FW) armor of future fusion reactors. However, a loss-of-coolant accident (LOCA) with simultaneous air ingress could lead to a temperature rise of the in-vessel components exceeding 1000°C, resulting in the formation and release of volatile and radioactive W oxides. In previous works it has been demonstrated that the addition of oxide forming alloying elements to pure tungsten results in the growth of a stable protective oxide scale. This prevents tungsten from oxidation at high temperatures, thus avoiding the potential risk of radioactive release during such a scenario. During normal operation, the surface of this self-passivating alloy will consist of pure tungsten, owing to preferential sputtering of the alloying elements by the plasma hydrogen isotopes. In this work, the development status of self-passivating W-Cr-Y alloys with different Cr and Y contents, manufactured by two different powder metallurgical routes, is reviewed. For both routes, the alloys exhibit a single metastable bcc phase with a dispersion of Y2O3 particles inhibiting grain growth. Isothermal oxidation tests in dry synthetic air at 1000°C for more than 100 h result in an oxidation rate three orders of magnitude lower than that of pure W due to the formation of a protective Cr2O3 layer. Oxidation tests at 1000°C under humid atmosphere reveal that Cr2O3 sublimation is enhanced while the formation of W-containing oxides is strongly reduced, evidencing that the material would withstand a LOCA for at least 10 days. However, at 1200°C the protection cannot be maintained. Besides oxidation, the mechanical properties and thermal conductivity are investigated. A single-phase W-10Cr-0.5Y alloy exhibits a high thermal shock resistance after 1000 ELM-like pulses at the JUDITH facility, where the alloy shows comparable performance to pure W. No indications of cracks are found after high heat flux tests up to 2 MW/m2 in the GLADIS facility. The microstructure and hardness remain unchanged after 3000 h at 650°C and 100 h at 700°C. W-Cr-Y alloys exposed to a pure deuterium plasma result in similar erosion yields than a simultaneously exposed reference W sample. First results on D retention are presented. The brazing feasibility of these alloys to EUROFER is studied using a Cu filler under different conditions, obtaining a low porosity and high continuity joint along both interfaces. Next steps in the development of this material are discussed including applicability of these alloys for the divertor, exposure to neutron irradiation and mock-up fabrication. | |
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Internacional
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Si |
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Nombre congreso
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ICFRM 16, 16th International Conference on Fusion Reactor Materials |
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Tipo de participación
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960 |
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Lugar del congreso
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La Jolla, California, USA |
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Revisores
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ISBN o ISSN
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0000000000 |
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DOI
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Fecha inicio congreso
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20/10/2019 |
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Fecha fin congreso
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26/10/2019 |
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Título de las actas
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16th International Conference on Fusion Reactor Materials |