Application of the FRI crack growth model for neutron-irradiated stainless steels in high-temperature water of a boiling water reactor environment

被引：12

作者：

Koshiishi, Masato ^{[1
]}

Hashimoto, Tsuneyuki ^{[1
]}

Obata, Ryoji ^{[2
]}

机构：

[1] Nippon Nucl Fuel Dev, 2163 Narita Cho, Oarai, Ibaraki 3111313, Japan

[2] Hitachi GE Nucl Energy, 1-1,Saiwai Cho 3 Chome, Hitachi, Ibaraki 3170073, Japan

来源：

CORROSION SCIENCE | 2017年 / 123卷

关键词：

Austenitic stainless steel; BWR; Neutron irradiation; Irradiation assisted stress corrosion cracking; Crack growth rate; Modelling; QUANTITATIVE PREDICTION; CORROSION BEHAVIOR;

D O I：

10.1016/j.corsci.2017.04.025

中图分类号：

T [工业技术];

学科分类号：

08 ;

摘要：

This study considered a methodology to reflect the effect on material properties by neutron irradiation onto the FRI model of the crack growth rate (CGR) of stress corrosion cracking developed at Tohoku University. Yield strength and strain hardening exponent were evaluated by a tensile test for irradiated stainless steels, and the relationship between mechanical property and strain to fracture of the oxide film was derived. Effect of the radiation-induced segregation on CGR was also discussed. The experimental tendencies for the CGR to increase with dose and almost saturate above 3 dpa were well replicated by the model calculations.

引用

页码：278 / 288

页数：11

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