VVER-1000型反应堆压力容器热老化分析评估  

Thermal Aging Assessment of RPV in VVER-1000 Reactor

作　　者：佟振峰[1] 崔贞北 赵继松[3] 张长义[1] 杨兴旺[3] 王克江[1] 刘维平[3] 杨文[1] 

机构地区：[1]中国原子能科学研究院反应堆工程研究设计所,北京102413 [2]中国核工业集团公司,北京100822 [3]江苏核电有限公司,江苏连云港222000

出　　处：《原子能科学技术》2015年第5期903-908,共6页Atomic Energy Science and Technology

基　　金：国家重点基础研究发展计划资助项目(2011CB610503);国家大型先进压水堆重大专项资助项目(2011ZX06004-002)

摘　　要：本文系统介绍了VVER-1000型反应堆压力容器(RPV)的温度监督情况,针对田湾核电站1#机组RPV的温度监督测试结果进行分析,评价运行3年后RPV力学性能(包括拉伸、冲击、断裂韧性)变化行为及热老化脆化机理,评估了当前田湾RPV服役运行后的热老化脆化状态和温度监督的时间安排。结果表明,温度监督样品经过堆内高温环境考验后,焊缝材料表现出一定程度的脆化特征,但母材、热影响区脆化不明显。与康采恩模型的结果和俄罗斯数据相比较后,认为田湾核电站1#机组RPV热老化脆化情况在合理范围内。VVER‐1000 reactor pressure vessel temperature surveillance is introduced systematically .The mechanical behaviors and thermal aging embrittlement mechanism of Tianwan unit 1 RPV after 3 years operation were analyzed and evaluated according to the results of temperature surveillance . The time scheme of next temperature surveillance was determined .The results show that after high temperature testing in reactor ,the sample weld for the temperature surveillance exhibits a certain degree of embrittlement ,but the embrittlement in the base metal and heat‐affected zone is not obvious .Comparing with Konzern model predicted value and the data from Russia ,it is considered that thermal aging embrittlement of RPV from Tianwan unit 1 is within a reasonable range .

关 键 词：VVER-1000 反应堆压力容器 热老化脆化 温度监督 

分 类 号：TL341[核科学技术—核技术及应用]