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In-pile tests for IASCC growth behavior of irradiated 316L stainless steel under simulated BWR condition in JMTR

JMTRにおけるBWR模擬条件下での照射済316Lステンレス鋼の照射誘起応力腐食割れ(IASCC)進展挙動に関する照射下試験

知見 康弘; 笠原 茂樹; 伊勢 英夫; 川口 佳彦*; 中野 純一; 西山 裕孝

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko*; Nakano, Junichi; Nishiyama, Yutaka

原子力機構では、沸騰水型軽水炉(BWR)炉内構造物の健全性評価の高度化の観点から、ステンレス鋼の応力腐食割れ(SCC)進展に及ぼす中性子/$$gamma$$線照射による材料及び水化学の変化の影響を評価するため、材料試験炉(JMTR)を用いた照射試験を計画している。炉内照射下及び炉外照射後試験でのSCC進展挙動及びその腐食電位(ECP)依存性の違いは、照射後試験データと比較可能な照射下試験データが非常に少ないため、完全には理解されていない。本報告では、過去の照射済ステンレス鋼のSCC進展データについて系統的に整理し、それを踏まえたJMTRにおけるBWR模擬条件下での照射済SUS316Lステンレス鋼のき裂進展挙動に関する照射下試験計画、及び照射下試験技術開発の概要を示す。

The Japan Atomic Energy Agency (JAEA) has a plan of irradiation tests using the Japan Materials Testing Reactor (JMTR), in order to evaluate the effects of change in material properties and water chemistry caused by the neutron/$$gamma$$-ray irradiation on stress corrosion crack (SCC) growth of stainless steels from the view points of the integrity of reactor core internals for boiling water reactors (BWRs). The difference of SCC growth and its electrochemical corrosion potential (ECP) dependence between in-pile and out-of-pile tests is not fully understood because of a few in-pile data which is comparable with out-of-pile database. This paper presents a systematic review on SCC growth data of irradiated stainless steels and the outline of the in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, together with the development of the in-pile test techniques.

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