TY - GEN
T1 - Quantifying the effects of straining-hardening and water chemistry on crack growth rates of 316L Ss welds in high temperature water
AU - Lu, Zhanpeng
AU - Sakaguchi, Kazuhiko
AU - Negishi, Koji
AU - Takeda, Yoichi
AU - Ito, Yuzuru
AU - Shoji, Tetsuo
PY - 2009/12/1
Y1 - 2009/12/1
N2 - A series of stress corrosion crack (SCC) growth rate tests with in-situ crack length monitoring for 316L stainless steel weld heat-affected zones with different Vickers hardness and 316 weld metals with different delta-ferrite contents in 288°C pure water with different concentrations of dissolved oxygen were performed. Typical intergranular stress corrosion cracking (IGSCC) and branching of the crack front were observed for two 316L HAZ specimens. Crack growth rate increases with Vickers Hardness in the 316L HAZ specimens. CGRs for 316L HAZ specimens in 2ppm and 7ppm DO are moderately higher than that in 0.2ppm DO water. After a single overloading, CGR tends to be lower for a period of time. Crack growth near the fusion line is monitored and the crack front of a 316L HAZ specimen is near the fusion line, while ACPD still shows an increase before stopping the test. SCC growth rates of two 316L weld metals with two ferrite contents, 9% in the crack growth region for high ferrite weld metal, and 6-7 % ferrite in the crack growth region were measured at 288°C in oxygenated pure water. Generally, CGRs in the high ferrite weld metal are higher than those in the low ferrite weld metal under the same test environment. The CGRs for the two weld metals in 2ppm DO water are higher than those in 200ppb DO water. For the weld metal with a low ferrite content, the CGR shows a tendency to become saturate at DO>2ppm. However, for the weld metal with a high ferrite content, the CGR in 2ppm DO water is higher than those in 7ppm and 11ppm DO water. Such an observation is confirmed by separate test.
AB - A series of stress corrosion crack (SCC) growth rate tests with in-situ crack length monitoring for 316L stainless steel weld heat-affected zones with different Vickers hardness and 316 weld metals with different delta-ferrite contents in 288°C pure water with different concentrations of dissolved oxygen were performed. Typical intergranular stress corrosion cracking (IGSCC) and branching of the crack front were observed for two 316L HAZ specimens. Crack growth rate increases with Vickers Hardness in the 316L HAZ specimens. CGRs for 316L HAZ specimens in 2ppm and 7ppm DO are moderately higher than that in 0.2ppm DO water. After a single overloading, CGR tends to be lower for a period of time. Crack growth near the fusion line is monitored and the crack front of a 316L HAZ specimen is near the fusion line, while ACPD still shows an increase before stopping the test. SCC growth rates of two 316L weld metals with two ferrite contents, 9% in the crack growth region for high ferrite weld metal, and 6-7 % ferrite in the crack growth region were measured at 288°C in oxygenated pure water. Generally, CGRs in the high ferrite weld metal are higher than those in the low ferrite weld metal under the same test environment. The CGRs for the two weld metals in 2ppm DO water are higher than those in 200ppb DO water. For the weld metal with a low ferrite content, the CGR shows a tendency to become saturate at DO>2ppm. However, for the weld metal with a high ferrite content, the CGR in 2ppm DO water is higher than those in 7ppm and 11ppm DO water. Such an observation is confirmed by separate test.
UR - http://www.scopus.com/inward/record.url?scp=78649361556&partnerID=8YFLogxK
UR - http://www.scopus.com/inward/citedby.url?scp=78649361556&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:78649361556
SN - 9781617388538
T3 - 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors 2009
SP - 636
EP - 645
BT - 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors 2009
T2 - 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors 2009
Y2 - 23 August 2009 through 27 August 2009
ER -