EVALUATION OF IRRADIATION ASSISTED STRESS-CORROSION CRACKING (IASCC) OF TYPE-316 STAINLESS-STEEL IRRADIATED IN FBR

Type 316 stainless steel from the core of the experimental fast breede r reactor (FBR) JOYO was examined by the slow strain rate tensile (SSR T) test in pure, oxygenated-water and air and by the electrochemical p otentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the ra diation-induced segregation (RIS). The solution annealed and 20% cold- worked materials had been irradiated at 425 degrees C to a neutron flu ence of 8.3 x 10(26) n/m(2) (> 0.1 MeV) which is equivalent to 40 disp lacement per atom (dpa). Intergranular cracking was induced by the SSR T in water at 200 and 300 degrees C, but was not observed on specimen tested in water at 60 degrees C and in air at 300 degrees C. This indi cates that irradiation increased a susceptibility to stress corrosion cracking (SCC) in water. After the EPR test, grain boundary etching wa s observed in addition to grain face etching. This suggests Cr depleti on may have occurred both at grain boundary and at defect clusters dur ing the irradiation. The results are compared with the behavior of sim ilar materials irradiated with different neutron spectrum.