SCC and Irradiation Properties of Metals under Supercritical-water Cooled Power Reactor Conditions

Corrosion issues in supercritical water reactor (SCWR) systems

“…From Ref. [100]. temperature in the subcritical regime followed by a decrease as temperature is increased into the supercritical regime.…”

“…Teysseyre and Was also determined the activation energy for crack growth to be about 105 kJ/mol for 304 and 85 kJ/mol for 316L [98]. Tsuchiya et al [100] measured the susceptibility to SCC of sensitized 304 tested in SCW with 8000 ppb dissolved oxygen as a percentage of IG cracking on the fracture surface. They observed a sharp decrease in %IG fraction on the fracture surface once they concluded that the upper limit of SCC susceptibility existed for sensitized 304SS at around 400°C and that sensitized 316 SS was not susceptible to IGSCC at 290°C and 550°C, but noticed the presence of transgranular SCC (TGSCC) at 550°C.…”

Effect of irradiation on stress corrosion cracking in supercritical water

“…Stress corrosion cracking (SCC) behavior of austenitic alloys in supercritical water has showed that many alloys (304, 316, 316L, 347) experience some degree of stress corrosion cracking, most of which is intergranular [1][2][3]. Most results show intergranular cracking on the gage surface even when intergranular cracking is not observed on the fracture surface.…”

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

“…The ductile fracture mode observed by Fournier is similar to that found in the 690 experiment in this study, with the exception of the region of "granular" fracture and the lack of necking in the sample. Tsuchiya et al [2003] investigated the behavior of sensitized 304 and 316 stainless steels using CERT experiments at 4x10 -7 s -1…”