PWR核电站蒸汽发生器传热管和主管道的应力腐蚀破裂研究

STRESS CORROSION CRACKING OF STEAM GENERATOR TUBE AND PRIMARY PIPE IN PWR TYPE NUCLEAR POWER PLANTS

用慢应变速率试验(SSRT)、恒载荷试验(CLT)和低周循环载荷试验方法研究以秦山和大亚湾核电站安全为目的的有关压力边界管道破裂始发事件应力腐蚀破裂(SCC)的行为,为评价管道的结构完整性提供支持性实验数据。研究的材料有核等级主管道焊接热影响区(WHAZ)316不锈钢(SS),核等级蒸汽发生器(SG)传热管材Incoloy-800、Inconel-600、Inconel-690和321SS。研究的影响因素包括材料冶金、表面喷丸处理、载荷、应变速率、循环载荷以及水化学条件对SCC的影响规律。

The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), con-stant load test (CLT) and low frequency cyclic loading test(LFCLT). The purpose of these tests is to getthe test data for evaluating the integrity of pressurized boundary of pipe in Qinshan and Guangdongnuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes inwelded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy--800, Inconel--600,Inconel--690 and 321 SS which are used for steam generator in PWR. The effects of material metallur-gy, shot--peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior ofSCC are investigated.