Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

• Published in: Acta materialia Vol. 274; p. 119059
• Main Authors: Haag, James V., Olszta, Matthew J., Edwards, Danny J., Jiang, Weilin, Setyawan, Wahyu
• Format: Journal Article
• Language: English
• Published: United States Elsevier Ltd 01.08.2024; Elsevier
• Subjects: Ion irradiation; Irradiation embrittlement; MATERIALS SCIENCE; Nuclear fusion; Precipitation; Transmission electron microscopy; Irradiation embrittlement; Precipitation; Ion irradiation; Transmission electron microscopy; Nuclear fusion
• ISSN: 1359-6454; 1873-2453
• DOI: 10.1016/j.actamat.2023.119059

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy's unique balance of high fracture toughness and high tungsten content; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. To alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni+ and He+ ion irradiations to mimic the expected displacement damage and He gas production expected after five years of service as materials for plasma facing components. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and a hexagonal W2C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation enhanced and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material's overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases. [Display omitted]