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Kaji, Yoshiyuki; Kondo, Keietsu; Aoyagi, Yoshiteru; Kato, Yoshiaki; Taguchi, Taketoshi; Takada, Fumiki; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Takakura, Kenichi*; et al.
Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1203 - 1216, 2011/08
In order to investigate the effect of neutron dose rate on tensile property and irradiation assisted stress corrosion cracking (IASCC) growth behavior, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating the temperature of boiling water reactor (BWR) up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the dose rate, but there was little effect on CGR. Increase of the yield strength of specimens irradiated with the low dose rate condition was caused by the increase of number density of frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, there was little effect on local plastic deformation behavior near crack tip in the crystal plasticity simulation.
Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko*; Nakano, Junichi; Nishiyama, Yutaka
Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1219 - 1228, 2011/08
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/
-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.
