EXPERIENCE AND ASSESSMENT OF STRESS-CORROSION CRACKING IN L-GRADE STAINLESS-STEEL BWR INTERNALS

Citation
Rm. Horn et al., EXPERIENCE AND ASSESSMENT OF STRESS-CORROSION CRACKING IN L-GRADE STAINLESS-STEEL BWR INTERNALS, Nuclear Engineering and Design, 174(3), 1997, pp. 313-325
Citations number
30
ISSN journal
00295493
Volume
174
Issue
3
Year of publication
1997
Pages
313 - 325
Database
ISI
SICI code
0029-5493(1997)174:3<313:EAAOSC>2.0.ZU;2-2
Abstract
The stress corrosion cracking (SCC) rate of reactor internals of boili ng water reactors (BWR) continues to increase with on-line operating y ears. The recent occurrences of cracking in the weld heat affected zon es of high carbon stainless steel core shrouds correlate with the year s of operation and the water chemistry history. Recently, cracking has also been found in shrouds that were constructed of low carbon or sta bilized stainless steels. While these steels are more resistant to int ergranular stress corrosion cracking (IGSCC) in the as-fabricated cond ition, this field experience establishes that the conditions under whi ch the materials will crack in core structures are attributable to the combined effects of high residual stresses, associated with the shrou d construction, the presence of a more aggressive, oxidizing environme nt in the core and to microstructural changes in the material. These c hanges result from the manufacturing process as well as thermal exposu re during operation. Studies of materials that have cracked in the fie ld, as well as high temperature material studies in the laboratory, ar e being performed to understand the mechanisms. The use of highly oxid izing, high purity water environments is integral to reproducing the c onditions for cracking. The status of the laboratory efforts to gain u nderstanding and to verify the mechanisms are presented. Modeling of I GSCC is also a key tool used to understand the cracking behavior of th e low carbon stainless steels. The PLEDGE (Plant Life Extension Diagno sis by GE) model is used to support the hypotheses that tie together t he role of the different contributing elements: residual stress, core water chemistry and microstructural features. The crack growth modelin g is also used to evaluate the benefits of different strategies to man age and mitigate cracking of reactor internals such as hydrogen water chemistry. (C) 1997 Published by Elsevier Science S.A.