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Pressurized Water Reactor Environment Effect on 316 Stainless Steel Stress Hardening/Softening: An Experimental Study

机译：压水堆环境对316不锈钢应力硬化/软化的影响：实验研究

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In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy's Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-Ⅰ) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-Ⅱ) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.
机译：在美国，大约有100座运行中的轻水反应堆（LWR），由压水反应堆（PWR）和沸水反应堆（BWR）组成。这些反应堆大多数是在1970年之前建造的，大多数反应堆的设计寿命是40年。预计到2030年，即使那些已获得美国核监管委员会（NRC）的20年延寿许可的反应堆也将开始达到其许可运行期的终点。出于经济原因，将许可证的使用期限延长60至80年是有益的，这将使现有工厂能够继续提供安全，清洁和经济的电力，而不会产生大量的温室气体排放。然而，环境疲劳是这些反应堆与老化相关的主要问题之一，并且可能在这些反应堆的长期可持续性方面造成障碍。为了解决一些与环境疲劳有关的问题，在能源部的轻水反应堆可持续性（LWRS）计划的赞助下，阿贡国家实验室（ANL）试图开发机械方法以更准确地估算轻水堆组件的寿命。在这种情况下，美国国家实验室对美国反应堆中广泛使用的316不锈钢（316SS）材料或类似等级的钢在不同的测试和环境条件下进行了许多疲劳实验。与传统的基于S〜N曲线的经验疲劳寿命估算方法相反，本DOE赞助的工作的目的是在不同的测试和环境条件下，更机械地了解材料的时效（例如，与时间有关的硬化和软化）。更好的机械理解将有助于开发基于计算机的高级建模工具，以更好地推断多轴应力状态下反应堆组件的应力-应变演变，从而有助于更准确地预测其疲劳寿命。本文（第一部分）讨论了在不同测试和环境条件下的316 SS疲劳实验以及相关的应力-应变结果。在另一篇论文（第二部分）中，讨论了相关的演化循环可塑性材料建模技术和结果。

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