原子能科学技术Issue(8)：1428-1433,6.DOI:10.7538/yzk.2015.49.08.1428
核安全一级主管道疲劳校核
Fatigue Check of Nuclear Safety Class 1 Reactor Coolant Pipe
王庆 1房永刚 1初起宝 1徐宇 1李海龙1
作者信息
- 1. 环境保护部核与辐射安全中心，北京 100082
- 折叠
摘要
Abstract
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-MKey words
nuclear safety Class 1 pipe/fatigue analysis/thermal ratcheting/ASM E/RCC-M分类
能源科技引用本文复制引用
王庆,房永刚,初起宝,徐宇,李海龙..核安全一级主管道疲劳校核[J].原子能科学技术,2015,(8):1428-1433,6.
