ASSESSMENT OF PRIMARY WATER-STRESS CORROSION CRACKING OF PWR STEAM-GENERATOR TUBES

被引：11

作者：

SHAH, VN ^{[1
]}

LOWENSTEIN, DB ^{[1
]}

TURNER, APL ^{[1
]}

WARD, SR ^{[1
]}

GORMAN, JA ^{[1
]}

MACDONALD, PE ^{[1
]}

WEIDENHAMER, GH ^{[1
]}

机构：

[1] NUCL REGULATORY COMMISS,WASHINGTON,DC 20555

来源：

NUCLEAR ENGINEERING AND DESIGN | 1992年 / 134卷 / 2-3期

关键词：

D O I：

10.1016/0029-5493(92)90139-M

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

This paper presents a model for assessing primary water stress corrosion cracking (PWSCC) damage to pressurized water reactor (PWR) steam generator tubes. PWSCC damage has been detected at three locations in the recirculating steam generator: expansion transition regions, U-bends, and tube dents. The model accounts for residual stresses, microstructure, and primary coolant temperatures. High residual tensile stresses and low mill-annealing temperatures contribute significantly to PWSCC. PWSCC is a thermally activated process; small increases in operating temperature accelerate damage. The model employs a Weibull distribution to represent tube degradation results obtained from inservice inspections and from destructive examinations. The model can be used to estimate the times to steam generator tube failures caused by PWSCC. This paper also evaluates remedial measures for PWSCC damage and presents a procedure for life assessment of steam generator tubes.

引用

页码：199 / 215

页数：17

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