Properties of unirradiated and irradiated Ti-6Al-4V alloy for ITER flexible connectors

摘要

The high strength (α + β) Ti-6Al-4V alloy was selected as the material for flexible attachments of the shield blanket modules in the ITER reactor. The different technologies used for manufacturing this alloy are: forging, stamping or pressing. The microstructures resulting from these processing methods can vary significantly and as a consequence the properties, including irradiation behavior, also vary. There are limited data available on the irradiation behavior of these materials. Specimens cut in the longitudinal and transversal directions of forged and stamped material were studied, with some of the specimens hydrogen charged to ～400 ppm H_2. In the unirradiated condition the forged alloy had slightly more ductility than the stamped alloy. The strength properties of both were practically the same. Neutron irradiation of these materials in the IVV-2M reactor reached doses of ～0.2 and 0.3 dpa at temperatures 240-260 °C. Irradiation resulted in substantial hardening and significant decrease of the fracture toughness of specimens from both materials.