A Study On Mechanical Behavior Of Austenitic Stainless Steels For Nuclear Reactors

Austenitic stainless steel is widely used in nuclear power plants because of its excellent corrosion, radiation and high temperature resistance and high strength. In the future the austenitic stainless steel is still important structural materials for generation IV reactor. 304 austenitic stainless steel is the most important one of the austenitic S.S. They are widely used in the primary and secondary loops of the existing generation II and III reactor. In both loops, austenitic S.S work at high temperature (over 300℃) and high pressure. So the research into the mechanical properties and high temperature mechanical properties of 304 austenitic stainless steel is important and meaningful. Because the comprehensive performance of 316L stainless steel is better than 304 stainless steel, 304 stainless steel will be replaced by 316L stainless steel as a new generation of nuclear power plants in a great head of road materials trends.In this study, rod of 304 and 316L austenitic stainless steels were heated to 1100℃for 60min, and then cooled with oil, water and air respectively. Effect of cooling method in mechanical properties of 304 and 316L stainless steels was examined through tensile test at temperatures ranging from room temperature to 400℃at strain rate 2×10-4s-1 .XRD was used to analyze the microstructure of the two candidate materials.Cooling method had no significant effect on mechanical properties of 304 and 316L austenitic stainless steel. Deformation-induced martensitic transformation in room-temperature tensile of 304 stainless steel caused stress hardening phenomenon. The middle of the fracture surface of 304 stainless steel is gray, like honeycom. There are many pits on the surface, consistent with the typical dimple characteristics of a typical plastic fracture.Dynamic strain aging (DSA)in 316L austenitic stainless steel was examined through tensile test at temperatures ranging from 300℃to 700℃at strain rate 2×10-4s-1 . Dynamic strain aging in 316L austenitic stainless steel did not accompany with a plateau of yield stress. 316L austenitic stainless steel exhibited both normal and inverse PLC effects at the temperatures and strain rate tested. The effective activation energy for serrated flow occurrence has been calculated to be about 254kJ/ mol. The dynamic strain aging caused by the interaction between substitutional solutes,such as Cr and moving dislocations is considered as the mechanism of serrated flow at the high temperatures in 316L stainless steel.