STRESS-CORROSION CRACKING OF ALLOY-600 USING THE CONSTANT STRAIN RATE TEST

被引：37

作者：

BULISCHECK, TS

VANROOYEN, D

机构：

来源：

CORROSION | 1981年 / 37卷 / 10期

关键词：

NUCLEAR REACTORS; PRESSURIZED WATER - Piping Systems;

D O I：

10.5006/1.3577544

中图分类号：

T [工业技术];

学科分类号：

08 ;

摘要：

Nuclear grade tubing was evaluated in high purity water at 365, 345, 325, and 290 C. Initial results indicate that a linear extrapolation of data from high temperature tests can be used to predict the service life of steam generator tubing that has been plastically deformed or is continually deforming by denting.

引用

页码：597 / 607

页数：11

共 8 条

[1] EFFECT OF CARBON CONTENT AND THERMAL-TREATMENT ON THE SCC BEHAVIOR OF INCONEL ALLOY 600 STEAM-GENERATOR TUBING [J].