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.

Water chemistry technology – one of the key technologies for safe and reliable nuclear power plant operation

“…In the SCC loop of the JMTR, IGSCC experiments under different water chemistry conditions were carried out but before sufficient amounts of data were obtained, the JMTR was stopped due to refurbishment of all reactor systems [73]. The refurbishment is almost finished and it is expected that new projects using the JMTR in-pile loops will be started soon after.…”

Experimental and numerical investigation of stress corrosion cracking of sensitized type 304 stainless steel under high-temperature and high-purity water

“…Andresen et al [26] evaluated crack growth of non-sensitized stainless steel in high-temperature water, and reported the effects of yield stress and corrosion potential on crack growth. Kaji et al [27] conducted in-core irradiation-assisted stress corrosion crack growth test using type 304 stainless steel in simulated BWR high-temperature water in Japan Materials Testing Reactor (JMTR) and evaluated the influence of neutron irradiation on crack growth rate. Lu et al [28] conducted SCC growth tests using welded specimens and pointed out that cracking behavior near a weld fusion line exhibited both macroscopic bifurcation and extensive microscopic branching, and crack growth near the weld fusion line was faster than that in a medially hardened region.…”

“…Many studies on crack initiation [24,25] and crack growth [26][27][28][29] have been carried out on materials exposed to high-temperature water, which is simulated environment in boiling water reactors (BWRs) and pressurized water reactors (PWRs). Andresen et al [26] evaluated crack growth of non-sensitized stainless steel in high-temperature water, and reported the effects of yield stress and corrosion potential on crack growth.…”