核安全一级主管道疲劳校核  

Fatigue Check of Nuclear Safety Class 1 Reactor Coolant Pipe

作　　者：王庆[1] 房永刚[1] 初起宝[1] 徐宇[1] 李海龙[1] 

机构地区：[1]环境保护部核与辐射安全中心,北京100082

出　　处：《原子能科学技术》2015年第8期1428-1433,共6页Atomic Energy Science and Technology

摘　　要：本文对某核电厂主管道疲劳及热棘轮进行了独立校核。校核采用基于RCC-M标准的ROCOCO软件,比较了RCC-M标准与ASME标准在核安全一级管道疲劳评价方面的差异。对比的主要方面包括疲劳设计的计算范围界定、一次加二次应力强度的计算方法、弹塑性修正系数的计算、动态载荷叠加方法等。通过对ROCOCO中与ASME标准不一致的算法进行修正,得到主管道冷段壁厚65 mm和55mm的疲劳使用系数和热棘轮设计裕量。结果表明:某核电厂主管道最小壁厚不能小于55 mm,55mm壁厚的热棘轮设计值达到许用值的95%。Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value.

关 键 词：核安全一级管道 疲劳分析 热棘轮 ASME RCC-M 

分 类 号：TL353[核科学技术—核技术及应用]