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Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 1025 n/m2 under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specime...

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing under a framework of cooperative research program between JAERI and the JAPC. In-pile IASCC growth tests have been successfully carried out usingmore » the compact tension (CT) type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 x 10{sup 25} n/m{sup 2} before the in-pile testing since 2004. The tests were carried out in pure water simulated boiling water reactor (BWR) coolant condition. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC. (authors)« less

In order to investigate the effect of neutron dose rate on tensile properly and irradiation stress corrosion cracking (IASCC) behavior, crack growth rate (CGR) and, tensile tests and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating boiling water reactor (BWR) environments up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). While radiation hardening increased with the dose rate, CGR was not affected by the dose rate. Increase of the yield strength of the low dose rate specimens 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, no dose rate effect on local plastic deformation behavior was found near crack tip in the crystal plasticity simulation