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On the effect of nuclear fission cladding stresses on Zirconium hydride orientation and dislocation strain energy fields via discrete dislocation plasticity and crystal plasticity finite element modelling
Journal of the Mechanics and Physics of Solids ( IF 5.0 ) Pub Date : 2024-11-02 , DOI: 10.1016/j.jmps.2024.105924
Christos Skamniotis, Daniel Long, Mark Wenman, Daniel S. Balint

The diffusion of hydrogen in Zircalloy fuel cladding components and its associated delayed hydride cracking (DHC) mechanism remain a bottleneck in nuclear fission. Through Crystal Plasticity Finite Element (CPFE) analysis at the grain scale (μm) and Discrete Dislocation Plasticity (DDP) at the hydride scale (nm), we explore how cladding stress history influences the dislocation network in a system of hydrides, and in turn, how this can impact hydrogen accumulation and embrittlement. CPFE indicates that high tensile stresses at service temperature can cause severe plasticity at a notch of a cladding component, leading to significant residual compressive stresses on service shutdown. As a result, hydrides evolve in this service scenario under a cyclic tensile-compressive background stress, which is found to enhance the ratchetting of dislocations compared to a typical constant background stress history and to eliminate the concentration of tensile residual hydrostatic stresses at the locations of dissolved hydrides. Since these tensile residual stresses drive the local accumulation of hydrogen during progressive precipitation-dissolution cycles, a key question is posed as to whether and how the sequencing of cladding stress-temperature reversals influences the growth rate of macro-hydride colonies. Simultaneously, we find that a large fraction of the total strain energy of hydrides is associated with the strain energy of dislocations and their interactions, posing the question of whether dislocation networks influence the energetically favourable hydride orientation. Our study provides a foundation for future studies of the DHC mechanism and drives the development of thermodynamically consistent dislocation-based models coupled with irradiation effects.

中文翻译:


通过离散位错塑性和晶体塑性有限元建模研究核裂变熔覆应力对氢化锆取向和位错应变能场的影响



氢在锆合金燃料包壳组件中的扩散及其相关的延迟氢化物裂解 (DHC) 机制仍然是核裂变的瓶颈。通过晶粒尺度 (μm) 的晶体塑性有限元 (CPFE) 和氢化物尺度 (nm) 的离散位错塑性 (DDP) 分析,我们探讨了熔覆应力历史如何影响氢化物系统中的位错网络,进而如何影响氢的积累和脆化。CPFE 表明,使用温度下的高拉伸应力会导致包层组件的缺口处出现严重的塑性,从而导致服务停机时产生明显的残余压应力。因此,氢化物在这种使用场景中在循环拉伸-压缩背景应力下演变,与典型的恒定背景应力历史相比,发现这增强了位错的棘轮化,并消除了拉伸残余静水应力在溶解氢化物位置的集中。由于这些拉伸残余应力在渐进的沉淀-溶解循环中驱动氢的局部积累,因此提出了一个关键问题,即包层应力-温度反转的顺序是否以及如何影响大氢化物菌落的生长速率。同时,我们发现氢化物总应变能的很大一部分与位错的应变能及其相互作用有关,这提出了位错网络是否影响能量上有利的氢化物取向的问题。我们的研究为 DHC 机制的未来研究提供了基础,并推动了基于热力学一致的位错耦合辐照效应模型的开发。
更新日期:2024-11-02
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