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首页> 外文会议> >Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach
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Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

机译：压水堆环境对316不锈钢材料参数的影响：基于循环塑性的演化材料建模方法

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摘要

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S～N) curves and Coffin-Manson type empirical relations. In most cases, the S～N curves are generated from uni-axial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S～N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.
机译：目前，核反应堆部件的疲劳寿命是根据经验方法估算的，如应力/应变对寿命（S〜N）曲线和科芬-曼森式经验关系式。在大多数情况下，S〜N曲线是根据单轴疲劳试验数据生成的，这些数据可能无法真正代表组件级的多轴应力状态。而且，S〜N曲线是基于试样的最终寿命，可能无法准确地表示材料行为的机械时变。这些差异导致疲劳寿命估算中存在很大的不确定性。我们提出了一种基于演化循环可塑性的建模方法，该方法可用于开发承受多轴应力状态的核反应堆组件的有限元模型。这些模型可用于更准确地预测反应堆组件随时间的应力应变变化，进而预测疲劳寿命。对在美国核反应堆中广泛使用的316不锈钢材料估算了模型参数。为不同的测试条件估算了模型参数，以了解它们随时间的变化以及它们对特定测试条件（如压水堆冷却剂条件）的敏感性。

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