Progress and Case Study on Probabilistic Assessment of Reactor Pressure Vessels under Pressurized Thermal Shock

Zengliang GAO
2015 Journal of Mechanical Engineering  
摘要:在压水堆核电站运行中,某些工况可能会使反应堆压力容器(Reactor pressure vessels, RPV)经受承压热冲击(Pressurized thermal shock, PTS)瞬态,这给含缺陷 RPV 的结构完整性带来了一定的挑战。简要介绍含缺陷 RPV 在 PTS 条件下的筛选准 则及其结构完整性评定方法,重点阐述 PTS 下含缺陷 RPV 的概率评定方法。概率评定方法采用概率断裂力学(Probabilistic fracture mechanics, PFM)分析,主要内容包括不确定因素统计分析(裂纹检出率、裂纹尺寸、材料性能等)、裂纹启裂模型及穿 透模型等。此外,还对适用于 PTS 分析的典型 PFM 程序进行评价。在此基础上,针对典型 RPV 利用自主开发的 PFM 程序 进行两个典型 PTS 瞬态的案例分析和结构完整性评定。分析结果表明在 60 年设计寿命内分析瞬态下该 RPV 的失效频率低 于核安全要求值。结合目前我国核电发展,针对 PTS 下 RPV 结构完整性概率评定提出几点建议。 关键词:承压热冲击;概率断裂力学;结构完整性评定;反应堆压力容器
more » ... 裂力学;结构完整性评定;反应堆压力容器 中图分类号:TL351 Abstract:A certain type of transients may cause the pressurized thermal shock(PTS) in the reactor pressure vessels(RPV) in pressurized water reactors(PWR) and may result in problem of structural integrity of RPV with flaws. The structural integrity assessment methods for RPV under PTS conditions are reviewed. The research progress and situation about these methods, especially probabilistic assessment, are introduced. Probabilistic assessment which uses the probabilistic fracture mechanics(PFM) analysis approach is reviewed, including uncertainty analysis (probability of detection, flaw characterization models, fracture property data, et al.), numerical calculating method on the failure probability, et al. Main PFM computational computer codes for PTS are evaluated. A PFM program is developed to calculate the failure probability. Two typical PTS transients for RPV and their structural integrities are analyzed. The results show that the failure probability for the RPV under these PTS transients is lower than the required value of nuclear safety for the design life of 60 years. Combined with Chinese nuclear power development, a few pieces of advices are proposed for probabilistic assessment of RPV structural integrity under PTS transients. Key words:pressurized thermal shock;probabilistic fracture mechanics;structural integrity assessment;reactor pressure vessels 0 前言 *
doi:10.3901/jme.2015.20.067 fatcat:dqb2idjpmnaf3gxwlo2hj3hzkq