Griffiths Malcolm
Department Mechanical & Materials Engineering, Queens University, Kingston, ON K7L 3N6, Canada.
Department of Mechanical & Aerospace Engineering, Carleton University, Ottawa, ON K1S 5B6, Canada.
Materials (Basel). 2021 May 17;14(10):2622. doi: 10.3390/ma14102622.
Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.
奥氏体不锈钢被用于钠冷快堆（SFR）和轻水堆（LWR）的堆芯内部结构，因为它们强度高，且在辐照后仍能保持韧性（在轻水堆中可达80 dpa），这与在低剂量（<1 dpa）下就会脆化的铁素体钢不同。对于快堆，内部结构的运行温度在400至550°C之间，燃料包壳的运行温度可达650°C。轻水堆的内部结构运行温度约在270至320°C之间，不过由于局部核加热，有些部件温度可能更高（超过400°C）。持续的可操作性依赖于能够理解和预测在长期运行过程中机械性能和尺寸稳定性如何变化。超过50年的试验堆辐照和动力堆运行经验积累了大量数据，据此可以评估辐照对奥氏体不锈钢性能的影响。本文将描述和讨论辐照对通过堆内和堆外（辐照后）测量及试验得出的固有机械性能（强度、延展性、韧性等）和尺寸稳定性的影响。将使用辐射损伤和气体产生模型评估主要观测结果。速率理论模型将用于展示辐照过程中的微观结构变化如何影响机械性能和尺寸稳定性。
