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.

在确定能够在聚变反应堆内部的不利环境中生存的材料的共同努力中，钨重合金已被提出作为候选材料。通过利用合金的高断裂韧性和高钨含量的独特平衡，实验试验和行为研究取得了采用它们的积极结果；然而，由于它们在聚变界的相对新颖性，人们对这些材料对反应堆内部长期高温辐照环境的反应仍缺乏了解。为了缓解这个问题并提供有关钨合金在模拟聚变环境中的行为的必要数据，对 90W-7Ni-3Fe 合金进行了高温顺序 Ni +^和He ⁺离子辐照以模拟预期的位移损伤和作为等离子体表面组件的材料使用五年后预期的氦气产量。原子级结构分析和纳米级化学作图已经确定了两种不同沉淀结构的形成，即表面局部 η-碳化物和六方 W ₂C 型碳化钨，两者似乎都起源于 W 和韧性相之间的双相界面。这种辐照增强和诱导沉淀物形成分别预计会通过双相界面的选择性脆化对这些材料产生不利影响，从而导致在长时间高温辐照期间材料的整体断裂韧性降低。据断言，在聚变反应堆运行服务期间任何可能接触 C 的情况都应尽量减少，以防止这些相的形成。