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Empirical equations of crack growth rates of neutron irradiated stainless steel under simulated BWR core conditions

中性子照射オーステナイト系ステンレス鋼の沸騰水型軽水炉炉心模擬環境下における亀裂進展速度に関する経験式の検討

笠原 茂樹; 福谷 耕司*; 知見 康弘; 藤井 克彦*; 端 邦樹

Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Hata, Kuniki

軽水炉の炉内構造物の構造健全性評価には、オーステナイト系ステンレス鋼への中性子照射影響を適切に反映したIASCC亀裂進展速度線図が必要である。本研究では、応力拡大係数Kと亀裂成長速度(da/dt)との間の関係式da/dt=M$$times$$K$$^{n}$$に基づき、BWR一次系模擬環境下で得られたIASCC亀裂進展速度データから各係数の最適化を行い、IASCC亀裂進展速度の実験式を構築した。Mとnは中性子照射量の関数として扱い、係数最適化ではBWR一次系でのNWCおよびHWCのそれぞれを模擬した条件下で得られたデータを用いた。構築した実験式とデータを比較し、実験式の 妥当性について検討した。

Disposition curves of crack growth rates (CGR) of stainless steel in appropriate consideration of IASCC are necessary for structural integrity assessment of reactor internals of BWRs. This paper describes empirical equations development of CGR (da/dt), as functions of stress intensity factors (K) and neutron dose (F) to contribute to improvement of the structural integrity assessment. Development started from a formula of da/dt=M$$times$$K$$^{n}$$, and on the assumption that "M" and "n" tend to be saturated with increasing F. Datasets for fitting were prepared consisting of CGR, F and K from the results of PIE under simulated NWC conditions. Data fitting with least square method was applied to the datasets to obtain the equation. The results from the empirical equation were compared with the measured crack growth data, and validity of the equations were discussed from the viewpoints of statistical analysis.

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