Evaluation of crack growth rates and microstructures near the crack tip of neutron-irradiated austenitic stainless steels in simulated BWR environment

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

https://doi.org/10.1007/978-3-319-68454-3_76

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.

Evaluation of crack growth rates and microstructures near crack tip of neutron-irradiated 316L stainless steels in simulated BWR environment

Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka; Seto, Hitoshi*; Chatani, Kazuhiro*; Kitsunai, Yuji*; Koshiishi, Masato*

no journal, , 

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests in simulated BWR water conditions (at 563 K) were performed using neutron-irradiated specimens made of 316L stainless steels, and the oxide film properties and locally deformed structures near the crack tip have been investigated by transmission electron microscopy (TEM). When electrochemical corrosion potential (ECP) of the materials was lowered by deaeration and hydrogen injection into feed water, apparent suppression of oxidation inside the cracks was observed as well as suppression of the crack growth rate (CGR). In the presentation, the TEM results of the locally deformed structures along the cracks are also reported, and the relation among the CGR, oxide film properties, and locally deformed structures is discussed.

Irradiation induced defects analysis in the heat affected zone of a reactor pressure vessel steel by transmission electron microscopy

Shimodaira, Masaki; Ha, Yoosung; Takamizawa, Hisashi; Seto, Hitoshi*; Katsuyama, Jinya

no journal, , 

no abstracts in English

The Neutron spectrometers in J-PARC

Kajimoto, Ryoichi; Yokoo, Tetsuya*; Nakamura, Mitsutaka; Shibata, Kaoru; Kawakita, Yukinobu; Matsuura, Masato*; Endo, Hitoshi*; Seto, Hideki*; Ito, Shinichi*; Nakajima, Kenji; et al.

no journal, , 

Effects of environmental mitigation and water radiolysis on crack growth in simulated BWR environment in highly irradiated 316L stainless steel

Chimi, Yasuhiro; Kasahara, Shigeki; Hata, Kuniki; Nishiyama, Yutaka; Seto, Hitoshi*; Chatani, Kazuhiro*; Kitsunai, Yuji*; Koshiishi, Masato*

no journal, , 

In order to investigate effects of environmental mitigation and water radiolysis caused by -rays from radioactive material on irradiation-assisted stress corrosion cracking (IASCC) growth behavior for highly irradiated material, crack growth tests in simulated BWR water conditions (at 563 K) are performed. The specimens made of 316L stainless steels are irradiated with neutrons up to 12 dpa in the Japan Materials Testing Reactor (JMTR). One of the specimens is annealed at 973 K for 1 hour to show almost recovered mechanical and micro-chemical properties corresponding to the unirradiated material. For low electrochemical corrosion potential (ECP) condition, the crack growth rate (CGR) is suppressed by about one order of magnitude in high stress intensity factor (K) condition. This result indicates that environmental mitigation for crack growth can be found even under severe conditions on material and stress factors. The effects of water radiolysis on the CGRs are discussed.

Relationship between crack growth rates and locally deformed structures in irradiated 316L stainless steels

Chimi, Yasuhiro; Kasahara, Shigeki; Nishiyama, Yutaka; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*

no journal, , 

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests using compact tension (CT) specimens made of neutron-irradiated 316L stainless steels (SSs) were performed in simulated BWR environments (at 288C). Moreover, microstructures of deformed areas were observed by transmission electron microscope (TEM) after straining tensile specimens made of neutron-irradiated 316L SSs. As a result, for lower neutron dose than 1.9 dpa, the crack growth rates (CGRs) show effective environmental mitigation and the deformed structures show tangling of dislocations. On the other hand, for higher neutron dose than 2.7 dpa, the CGRs show small environmental mitigation and the deformed structures consist mainly of dislocation channels. From the relationship between CGRs and deformed structures, mechanisms on IASCC growth will be discussed.