TEM Characterization of Irradiation-induced Dislocation Loops and Voids in ion-irradiated Pure Chromium
TEM Characterization of Irradiation-induced Dislocation Loops and Voids in ion-irradiated Pure Chromium
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DOI：
10.1016/j.jnucmat.2022.153920
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发表时间：
2022-07
影响因子：
3.1
通讯作者：
L. Cui;H.L. Yang;Y. F. Du;Q. Shi;S. Kano;H. Abe
中科院分区：
文献类型：
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作者：
L. Cui;H.L. Yang;Y. F. Du;Q. Shi;S. Kano;H. Abe
铬（Cr）基涂层已被提出用于锆合金燃料棒，以提高水冷反应堆的事故容限。然而，铬对辐射的反应在以前的研究中受到了有限的关注。为了更好地了解Cr的辐射损伤，为未来的核应用，在本研究中，辐照诱导的位错环和空洞的微观结构的演变进行了研究，在纯Cr受到2.8 MeV的铁离子辐照在550 °C（属于第III阶段，中温）。TEM观察结果表明，位错环的分布将SRIM模拟的辐照深度扩展到3.5 µm，而空洞仅在辐照区域形成，深度为1.3 µm。提出了一种综合考虑边视和内外分析的严格的确定习惯面的方法。用这种方法，通常被认为是不安全的区域变成了确定位错环性质的有利区域。确定位错环的性质，包括间隙和空位类型，<111>在所有区域中具有1/2 Burgers矢量。在辐照区内，回线接近半间隙半空位，而辐照区以外的间隙回线成为主导。通过空隙尺寸和密度来量化辐射诱导的溶胀。结果表明，该模型与SRIM模拟的辐照剂量分布基本吻合。接近峰值区域的溶胀率为约0.3%/dpa。结果表明，Cr倾向于经历一些相似的辐照循环演变的W基，Mo基材料，因为它们是在周期表中的同一组，尽管在循环的特性，如辐照温度依赖的位错环的形成的差异，以及观察。本研究不仅对补充铬辐照损伤的基础知识，而且对铬涂层在事故容限燃料包壳或其它核结构材料中的应用具有重要意义。
Chromium (Cr) based coatings have been proposed for Zircaloy fuel rods to enhance the accident tolerance in water-cooled reactors. However, the response of Cr to irradiation has received limited attention in prior studies. To get a better understanding of radiation damage in Cr for future nuclear applications, in the present study, the irradiation-induced microstructure evolution of dislocation loops and voids was studied in pure Cr subjected to 2.8 MeV iron ions irradiation at 550 °C (belonging to stage III, intermediate-temperature). TEM observation results show that the distribution of dislocation loops extends the SRIM simulated irradiation depth to 3.5 µm, while voids were only formed in the irradiated region up to a depth of 1.3 µm. A rigorous method with combined consideration of edge-on view and inside-outside analysis was proposed to determine the habit plane. With this method, what has normally considered as an unsafe zone becomes a favored zone for determining the nature of dislocation loops. The nature of dislocation loops was determined including both interstitial and vacancy types with ½<111> Burgers vector in all regions. Within the irradiation region, the loops are approaching half interstitial and half vacancy, while beyond the irradiated region interstitial loops became dominant. The irradiation-induced swelling was quantified by the void size and density. The swellings are mostly corresponding well with the irradiation dose distribution which is simulated by SRIM. The swelling rate close to the peak region is around 0.3%/dpa. Results demonstrate that Cr tends to experience some similarities in irradiation loop evolution to W-based, Mo-based materials, as they are in the same group in the periodic table, despite the discrepancy in the characteristic of loops such as the irradiation temperature-dependent dislocation loop formation was observed as well. The present study is significant not only for supplementing the fundamental knowledge of irradiation damage in Cr but also for the application of Cr coating to the accident tolerant fuel claddings or other nuclear structure materials.