Influence of self-interstitial mobility on damage accumulation in zirconium under fission irradiation conditions
Author:
Publisher
Elsevier BV
Subject
Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics
Reference12 articles.
1. Statistics of primary damage creation in high-energy displacement cascades in copper and zirconium
2. Identification and morphology of point defect clusters created in displacement cascades in α-zirconium
3. Defect production, annealing kinetics and damage evolution in α-Fe: An atomic-scale computer simulation
4. Simulation of radiation damage in Fe alloys: an object kinetic Monte Carlo approach
Cited by 8 articles. 订阅此论文施引文献 订阅此论文施引文献,注册后可以免费订阅5篇论文的施引文献,订阅后可以查看论文全部施引文献
1. Interactions between clusters of self-interstitial atoms via a conservative climb in BCC–Fe;Philosophical Magazine;2018-06-21
2. Effect of ion irradiation of the metal matrix on the oxidation rate of Zircaloy-4;Corrosion Science;2018-05
3. Temperature dependence of migration features of self-interstitials in zirconium;Chinese Physics B;2017-12
4. The Damage Cascade;Fundamentals of Radiation Materials Science;2016-07-09
5. Kinetic Monte Carlo Simulations of Irradiation Effects;Comprehensive Nuclear Materials;2012
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