沸騰水型軽水炉炉内構造物用低炭素含有オーステナイト系ステンレス鋼の溶接熱影響部を対象とした中性子照射データの調査(受託研究)

Data investigation on effects of neutron irradiation on material properties of the heat-affected zone near weld fusion line of austenitic stainless steel with low carbon for core internals of boiling water reactors (Contract Research)

笠原 茂樹; 端 邦樹 ; 岩田 景子 ; 知見 康弘

Kasahara, Shigeki; Hata, Kuniki; Iwata, Keiko; Chimi, Yasuhiro

2000年代初頭以降、国内の発電用沸騰水型軽水炉の一次系冷却材環境において非鋭敏化低炭素ステンレス鋼の溶接部近傍に粒界型応力腐食割れによる損傷が顕在化したことを受け、メカニズム解明研究と対策技術開発が進められている。これまでの調査、検討では、低炭素ステンレス鋼の溶接部近傍は、溶接入熱による膨張と収縮によって局所ひずみが蓄積して硬さが上昇したことが粒界型応力腐食割れの材料因子となっていると考えられており、硬さ上昇と粒界型応力腐食割れの因果関係の解明が急務となっている。上記に加えて、沸騰水型軽水炉の炉内構造物の健全性評価に当たっては、溶接熱影響部の中性子照射による硬さの上昇(照射硬化)の重畳を考慮した粒界型応力腐食割れの評価が求められ、溶接熱影響硬化部に対する照射影響評価に資する多角的、系統的なデータの拡充と公開が望まれる。本調査では、これまで未公開だったデータを中心に、原子力安全基盤機構が実施した「低炭素ステンレス鋼応力腐食割れ進展への中性子照射影響実証」事業で取得された低炭素ステンレス鋼溶接熱影響部の機械的性質や高温水環境中亀裂進展速度等に係る照射データを調査、収集した。

It has been reported that intergranular stress corrosion cracking (IGSCC) has occurred near weld fusion lines of low-carbon, namely unsensitized austenitic stainless steel for pipings and reactor internals used under the primary coolant environment of boiling water reactors in Japan since the early 2000s. It becomes one of the critical technical issues clarification of the mechanism and development of countermeasure techniques for IGSCC of low-carbon-containing stainless steel. From previous research, the hardness of stainless steel is increased due to the accumulation of local strain, after expansion and contraction during welding heat input, and the increment of hardness in such heat-affected zone is recognized one of the possible material factors caused by IGSCC. In particular, for boiling water reactor internal structures, it is essential to evaluate IGSCC taking into account neutron irradiation as well as strain accumulation for hardness increase, and it is desirable to accumulate multifaceted and systematic data that can be dedicated to evaluating the irradiation effect of weld heat-affected and hardened zones. In this study, we investigated and collected irradiation data on the crack growth rates and other material properties evaluated under simulated primary water conditions of boiling water reactor environments for neutron-irradiated low-carbon stainless steel weld heat-affected zones. Those were obtained through the ENI project of the Japan Nuclear Energy Safety Organization, including data that had not been made public until now.