Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

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

“…Use of FBR-irradiated materials [211,212] and irradiation in FBRs at temperatures lower than 370 • C [213] have been tried to study the IASCC behavior at higher doses. Recent SSRT test results in PWR water conditions showed that the IASCC susceptibility of FBRirradiated type 316 SSs was much lower than PWRirradiated data [213].…”

A systematic approach for the study of radiation-induced segregation/depletion at grain boundaries in steels

“…The annihilation of point defects at grain boundary in particular creates non-equilibrium solute profiles that would not occur under thermal conditions. Consequently, radiation-induced segregation (RIS) or depletion (RID) is a particular important phenomenon that has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials [1][2][3][4]. Okamoto and Wiedersich [5] suggested that depletion or segregation of a solute element to a grain boundary depends on its diffusion mechanism and therefore in a first approximation on its size.…”

Effects of water and irradiation temperatures on IASCC susceptibility of type 316 stainless steel

“…7 summarizes the effect of water temperature and irradiation temperature on IASCC susceptibility of type 316(L) stainless steels. Data from previous studies [1][2][3] is also plotted in the figure. In oxygenated water below 513 K, specimens irradiated below 473 K shows no IASCC susceptibility.…”

“…Type 316 stainless steel irradiated to 7.4 dpa at 333 and 473 K showed no IASCC susceptibility in oxygenated water at 333 and 473 K, respectively [2], while the stainless steel irradiated at 698 K to 40 dpa showed IASCC susceptibility in 473 K water [1]. In oxygenated water around 573 K, however, IASCC susceptibility was exhibited after the irradiation even to 1 dpa at 333 K [5].…”

“…As a result, austenitic stainless steels lose ductility and corrosion resistance at/near grain boundaries, and become sensitive to irradiation assisted stress corrosion cracking (IASCC), which has been studied extensively in the field of light water reactors. However, there are few studies of IASCC susceptibility in relatively low temperature water like the ITER operating condition [1][2][3][4]. During start-up and shut-down of the reactor, plasma disruptions and bake-out operation of ITER, the structural materials will be heated to temperatures that will be different from normal operating temperatures.…”