摘要
由于高的热效率和简单的系统组成,超临界水堆(SCWR)被认为是第四代核反应堆的一种选择。超临界水堆的关键问题之一是核心部件尤其是燃料组件包壳的材料。这些材料在高温下的力学性能、腐蚀和应力腐蚀开裂敏感性以及抗辐射性能等对核电厂的安全运行至关重要。本文对SCWR包壳候选材料的F/M类材料P92钢进行了高温低周疲劳实验研究。实验温度为600和650℃,控制方式为总应变控制,应变范围均为±0.2%～±0.6%。实验结果表明,在两种温度下,P92钢均为循环软化材料,但未出现循环稳定现象。由于温度升高,塑性增强,P92钢在650℃下的宏观裂纹出现周次比率随应变范围的增加,下降比较平缓,且650℃下的失效寿命显著高于600℃下的失效寿命。并得到了两种温度下的稳定循环应力-塑性应变的关系以及循环失效寿命和应变的关系。
A supercritical water cooled reactor（SCWR）is being considered as a candidate reactor of the Generation Ⅳ nuclear reactors due to its high thermal efficiency and simple system composition.A critical question to attain is to choose proper materials for the core components,especially for the fuel cladding.The mechanic properties,corrosion and stress corrosion cracking susceptibilities,radiation resistances,etc.,of these materials at high temperature are extremely important for the safety of nuclear power plant.The paper presents the low cycle fatigue behaviors of P92,a kind of F/M type candidate materials for the SCWR.The experiments were carried out at 600 ℃ and 650℃ with total strain controlled.The strain range is from ±0.2%-±0.6%,respectively.The results show that P92steel is cyclic strain softening at both temperatures, but stable cyclic phenomena were not observed.The decline ratio of macro-crack appearance with the strain range increasing is milder at 650℃ than that at 600℃,and the cycles to failure are remarkably higher at 650℃ than those at 600℃ under the same total strain ranges.The relationship of cycle stable stress vs.strain range and number of cycles to failure vs.total strain range were obtained.
出处
《原子能科学技术》
EI
CAS
CSCD
北大核心
2010年第10期1212-1216,共5页
Atomic Energy Science and Technology
基金
国家重点基础研究发展计划资助项目(2007CB209803973)
长江学者和创新团队发展计划资助项目(IRT0720)
关键词
P92钢
F/M钢
超临界水堆
疲劳
P92steel
F/M steel
supercritical water cooled reactor
fatigue
