核融合模擬環境におけるタングステン系材料の特性評価(Characterizing Tungsten-Based Materials in a Simulated Fusion Environment)

2024-07-13

2024-07-12 パシフィック・ノースウェスト国立研究所(PNNL)

核融合炉内部は非常に過酷な環境であり、高温や強い放射線に耐えられる材料は少ないです。タングステン重合金（WHA）は、高い融点と熱伝導率を持つタングステンに、延性と破壊靭性を加えた有望な材料です。研究者たちは、5年の使用後を模倣した高温とイオン照射環境にW-Ni-Fe WHAを曝露し、原子レベルでの構造解析を行いました。その結果、炭素不純物による選択的脆化が観察されました。この研究は、核融合炉内でのWHAの挙動を理解し、長期間の高温照射環境での耐久性を評価する基盤となります。

＜関連情報＞

重イオン照射されたタングステン重合金における核融合模擬環境下での境界析出の特性評価 Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

James V. Haag IV, Matthew J. Olszta, Danny J. Edwards, Weilin Jiang, Wahyu Setyawan
Acta Materialia Available online 2 June 2023
DOI:https://doi.org/10.1016/j.actamat.2023.119059

Abstract

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.