国产反应堆压力容器的辐照脆化行为及预测  

Irradiation Embrittlement Behavior and Prediction of Domestic Reactor Pressure Vessel

作　　者：林虎[1] 钟巍华[1] 佟振峰[1,2] 宁广胜 张长义[1] 杨文[1] LIN Hu;ZHONG Weihua;TONG Zhenfeng;NING Guangsheng;ZHANG Changyi;YANG Wen(China Institute of Atomic Energy,Beijing 102413,China;North China Electric Power University,Beijing 102206,China)

机构地区：[1]中国原子能科学研究院,北京102413 [2]华北电力大学,北京102206

出　　处：《原子能科学技术》2021年第7期1170-1176,共7页Atomic Energy Science and Technology

基　　金：大型先进压水堆核电站重大专项(2018ZX06002008)。

摘　　要：反应堆压力容器(RPV)是保障核电站运行安全性、经济性的核心构件。对RPV的完整性评估而言辐照脆化是必须面对的问题。我国已开发了第三代设计寿命为60 a的核电站。当达到寿期末时,辐照脆化的行为是未知的,这给国产RPV的辐照脆化预测带来了困难。为研究高注量下的辐照脆化行为,对A508-3钢的材料力学性能试样进行辐照考验,辐照温度为(288±8)℃,中子注量水平达到反应堆压力容器60 a寿期末的辐照水平1×10^(20) cm^(-2);开展拉伸、冲击和断裂韧性试验,分析辐照脆化行为,在EONY模型基础上,提出针对国产RPV钢的改进的辐照脆化模型。模型的有效性被试验数据证实,其可准确预测国内A508-3材料的辐照脆化趋势。Reactor pressure vessel(RPV)is the core component that related to operation safety and economy of nuclear power plants(NPPs).Irradiation embrittlement is the most important factor in RPV structural integrity evaluation.China is now developing the 3 rd generation RPV with 60 years’designed life time.However the irradiation embrittlement behavior and prediction model for the neutron fluence equal to 60 effective full power year(EFPY)are unknown,which may lead to difficulty in predicting irradiation embrittlement of domestic RPV.In this work,mechanical test specimens of A508-3 steel were irradiated at temperature of(288±8)℃and the fluence is 1×10^(20) cm^(-2)(E>1 MeV)which is equivalent to 60 EFPY.Then tensile,impact and fracture toughness tests were carried out,and the irradiation embrittlement behavior was analyzed.On the basis of EONY,a modified irradiation embrittlement prediction model for domestic RPV steel was established.Then the model was verified by the test data.The result shows that the prediction model can predict the irradiation embrittlement of domestic A508-3 steel accurately and reliably.

关 键 词：反应堆压力容器 A508-3钢 中子辐照 辐照脆化 预测模型 

分 类 号：TL341[核科学技术—核技术及应用]