低铜压力容器钢辐照脆化效应实验研究  

EXPERIMENTAL INVESTIGATION OF RADIATION EMBRITTLEMENT EFFECTS OF LOW COPPER REACTOR PRESSURE VESSEL STEELS

作　　者：徐远超[1] 贾学军[1] 张长义[1] 杨继材[1] 杨文斗[1] 

机构地区：[1]中国原子能科学研究院

出　　处：《核科学与工程》1998年第2期146-152,共7页Nuclear Science and Engineering

摘　　要：研究了快中子注量率、注量和辐照温度等辐照参数对低铜压力容器钢的辐照脆化程度的影响，从而将实验堆辐照试验数据与动力堆监督试验数据关联。采用了仪表化冲击试验设备和双曲正切函数回归计算的数据处理方法，因而确保了实验结果的准确性。应用半经验公式将仪表化冲击试验数据转化为动态断裂韧性。为压力容器使用寿命评估和新建核电站压力容器设计提供了材料辐照脆化数据。The radiation embrittlement of two types of low copper reactor pressure vessel steels is studied. The relation between the material damage embrittlement extent and irradiation parameters such as fast neutron flux、 fluence and temperature is investigated, so with the test data of SA508 3 steel irradiated in the test reactor to be correlated to PWR pressure vessel surveillance test data. The advanced instrumented impact test equipment and the method for treating test data with the hyperbolic tangent function ensure the accuracy of results. Instrumented charpy V notch data was converted into dynamic fracture toughness by means of semi empiric formula. This investigation provides reliable data on the material irradiation embrittlement for predicting lifetime of pressure vessels in operation or in design.

关 键 词：压力容器 钢 辐照参数 脆化 断裂韧性 反应堆 

分 类 号：TL341[核科学技术—核技术及应用]