In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR

Abstract

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 Â 10 25 n/m 2 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 specimens, the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate was considered to be small, because the in-core data under the same ECP condition were similar to the ex-core data under the DO ¼ 32 ppm condition.

Current understanding of radiation-induced degradation in light water reactor structural materials

“…One unique CGR data set on sensitized 304 SSs measured in an operating BWR showed that much higher CGRs were observed at an incore position than at an out-of-core position and these corresponded to higher ECP for the in-core position [207]. In-core tests (in-pile or in-reactor tests) of SCC growth of irradiated SSs have been conducted in the Halden reactor [202,204] and the Japan Material Testing Reactor [208]. Figure 16 compares CGR data of 304 SSs from in-pile tests and PIEs in BWR NWC water conditions [208].…”

“…In-core tests (in-pile or in-reactor tests) of SCC growth of irradiated SSs have been conducted in the Halden reactor [202,204] and the Japan Material Testing Reactor [208]. Figure 16 compares CGR data of 304 SSs from in-pile tests and PIEs in BWR NWC water conditions [208]. Comparison of the in-core data with corresponding PIE data (ex-core, out-of-core or out-ofpile data) showed that the in-core CGRs were similar to or slightly higher than PIE data at high ECP in BWR water conditions.…”

Failure Behavior of Sensitized SUS304 Stainless Steel with Small Surface Pre-Crack under Constant Load Condition in Simulated BWR Environment

“…Due to lifetime extension of BWR, since integrity evaluation is one of the most serious problems to be taken into consideration, the behavior of these existing stress corrosion cracks, regardless of small one or long one, must be clarified to guarantee the safety of the system. So far, many researchers have studied the stress corrosion cracking behavior of long crack in austenitic stainless steels under simulated BWR environment using CT specimen together with AC or DC potential drop techniques for crack length measurement (5)- (8) and proposed models to explain SCC behavior (9)- (11) . Furthermore, relationship between crack growth rate (da/dt) and stress intensity factor (K) has been successfully obtained and compared with the existing reference curves, such as the JSME code for Nuclear Power Generation Plants (12) , the NRC NUREG-0313 Rev.2 (13) and the Swedish Nuclear Power Inspectorate's Regulations SKIFS (14) .…”