ZrO2-corrosion-layers and their grain boundary networks

锆合金用于水冷反应堆的核燃料包壳。这些合金在接触当地环境，如热水和蒸汽时会被腐蚀(Féroon 2012)。最初的氧化会迅速导致氧化锆层的形成，从而降低氧化速度。带电物种在氧化层中的迁移在很大程度上受微观结构的控制(Garner等人。2015、2017)和氧化层的晶界网络(Gertsman等人)。1997)。最初是靠近金属-氧化物界面的非平衡四方相。然而，随着氧化层厚度的增加，金属与氧化物之间的反应界面位置离表面更远。氧化层中的压应力随着距离金属/氧化物界面的距离而减小，最终不足以稳定四方相。在这个具有挑战性的项目中，我们的目标是使用最先进的介观到纳米尺度的表征技术来表征不同晶界网络的行为，包括透射电子显微镜和电子背散射衍射(EBSD)技术。我们将通过实验改变锆合金，并研究这如何影响由此产生的锆氧化层的形成和微观结构特性。最终，实验数据将被用来使用不同的建模方法来探索微观结构。参考文献：Féron，D.(2012)序言。《核腐蚀科学与工程》pp。XXVII-xxix。[2]Elsevier.Garner，A.，Hu，J.，Harte，A.，Frankel，P.，Grovenor，C.，Lozano-Perez，S.和Preuss，M.(2015).锡含量对锆合金中氧化物织构和微观结构形成的影响.材料学报，99,259-272.加纳，A.，Frankel，P.，Partezana，J.和Preuss，M.(2017).衬底织构和氧化温度对锆合金中氧化物织构发展的影响.核材料杂志，484,347-356.格茨曼，v.Y.，Zilyaev，A.P.和Szpuar，J.A.(1997)单斜氧化锆晶界取向不均分布.材料科学与工程中的建模与模拟，5，35-52。

Zr-alloys are used for nuclear fuel cladding in water-cooled reactors. These alloys corrode in contact with their local-environment, e.g. hot water and steam (Féron 2012). Initially oxidation rapidly results in the formation of a zirconium oxide layer that reduces the oxidation rate. The migration of charged species in the oxide layer is largely controlled by the microstructure (Garner et al. 2015, 2017) and grain boundary network of the oxide layer (Gertsman et al. 1997).Initially a non-equilibrium tetragonal phase close to the metal-oxide interface. However as the Oxide layer gains thickness the location of the reaction interface between metal and oxide are further from the surface. The compressive stress in the oxide layer decrease with distance from the metal/oxide interface to eventually be insufficient to stabilise the tetragonal phase. It transforms to monoclinic ZrO2, which is the stable phase at atmospheric pressure and temperature below 1500 K. In this challenging project, we aim at characterizing the behaviour of different grain boundary networks using state-of-the-art meso- to nano-scale characterisation techniques including transmission electron microscopy (TEM) and electron backscatter diffraction (EBSD) techniques. We will experimentally alter Zr-alloys and investigate how this affects the formation and microstructural properties of the resulting Zr-oxide layers. Eventually the experimental data will be used to explore the microstructures using different modelling approaches.References:Féron, D. (2012) Preface. In Nuclear Corrosion Science and Engineering pp. xxvii-xxix. Elsevier.Garner, A., Hu, J., Harte, A., Frankel, P., Grovenor, C., Lozano-Perez, S., and Preuss, M. (2015) The effect of Sn concentration on oxide texture and microstructure formation in zirconium alloys. Acta Materialia, 99, 259-272.Garner, A., Frankel, P., Partezana, J., and Preuss, M. (2017) The effect of substrate texture and oxidation temperature on oxide texture development in zirconium alloys. Journal of Nuclear Materials, 484, 347-356.Gertsman, V.Y., Zhilyaev, A.P., and Szpunar, J.A. (1997) Grain boundary misorientation distributions in monoclinic zirconia. Modelling and Simulation in Materials Science and Engineering, 5, 35-52.