Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

被引：9
作者：
Rao, Appajosula S. [1 ]
机构：
[1] US Nucl Regulatory Commiss, Off Nucl Regulatory Res, Div Engn, Washington, DC 20555 USA
来源：
NUCLEAR ENGINEERING AND DESIGN | 2014年 / 269卷
关键词：
D O I：
10.1016/j.nucengdes.2013.08.010
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：
Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies. Published by Elsevier B.V.
引用
收藏
页码：78 / 82
页数：5
相关论文
共 17 条
[1]  
Allen T. R., 2007, P 13 INT C ENV DEGR
[2]  
Brown K. S., 1987, P 3 INT S ENV DEGR M, P243
[3]   Radiation-induced material changes and susceptibility to intergranular failure of light-water-reactor core internals [J].
Bruemmer, SM ;
Simonen, EP ;
Scott, PM ;
Andresen, PL ;
Was, GS ;
Nelson, JL .
JOURNAL OF NUCLEAR MATERIALS, 1999, 274 (03) :299-314
[4]  
Byrne S., 2001, P 10 INT C ENV DEGR
[5]   A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels [J].
Chopra, O. K. ;
Rao, A. S. .
JOURNAL OF NUCLEAR MATERIALS, 2011, 409 (03) :235-256
[6]  
Chopra O. K., 2008, NUREGCR6960
[7]  
CHUNG HM, 2006, NUREGCR6892
[8]  
Edwards D., 2005, P 12 INT C ENV DEGR, P419
[9]  
Edwards D. J., 2007, P 13 INT C ENV DEGR
[10]  
Electric Power Research Institute, 2004, 1011127 EPRI
12