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Aoki, So; Kondo, Keietsu; Kaji, Yoshiyuki; Yamamoto, Masahiro
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.663 - 672, 2018/00
Times Cited Count：2 Percentile：58(Materials Science, Multidisciplinary)The objective of this study was to clarify effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L steels. Both steels were cold rolled to 20% thickness reduction (CW) and then followed by long-term thermal aging at 288
C for 14,000 h (LTA). Crevice Bent Beam (CBB) test was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The present results of the CBB tests showed that 304L CW + LTA exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of 316L increased by the combination of cold rolling and long-term thermal aging. To understand these results, evaluation on the changes in microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment has been developed, and their correlation with the SCC initiation susceptibility is discussed.
Soma, Yasutaka; Kato, Chiaki; Ueno, Fumiyoshi
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.509 - 521, 2018/00
In-situ electrochemical measurement within crevice of stainless steel in 288C water has been conducted to analyze crevice water chemistry. Small sensors (
250
m) measured local solution electrical conductivity, 
, polarization resistance, and electrochemical corrosion potential. Real-time response of the as functions of bulk water conductivity, dissolved oxygen (DO) concentration has been quantitatively analyzed. The effect of geometrical factors on the crevice environment was also studied. The differ more than an order of magnitude depending on the oxygen potential inside the crevice. The increased by small amount of bulk DO (e.g. 30 ppb). Maximum was observed with DO of 32000 ppb and became more than 100 times higher than that of bulk water. Crevice geometry affected significantly on the water chemistry inside.
Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00
Times Cited Count：2 Percentile：58(Materials Science, Multidisciplinary)In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at 288C on neutron-irradiated 316L stainless steels (SSs) at 12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at 
2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.
