Constant extension rate tensile (CERT) specimens were irradiated in the core of a commercial operating BWR. Subsequently to irradiation, CERT testing was performed in a test loop attached to the reactor water clean-up system in the same BWR, using reactor water at a high flow rate. Testing was performed in BWR normal water chemistry (NWC), and BWR hydrogen water chemistry (HWC). Type 316 SS was much less susceptible to IASCC than Type 304 SS in both NWC and HWC. In addition, high purity (HP) Type 304 SS was more susceptible to IASCC than commercial purity (CP) Type 304 SS in NWC at intermediate fluence (~1.1021 n/cm2, E>l MeV). The results suggest that chromium depletion due to radiation induced segregation (RIS) is an operating mechanism in NWC, and the only one of? importance at fluences below 3.1021 n/cm2 (E>l MeV). At higher fluence another mechanism supervenes. Detailed microchemical analysis was conducted on unirradiated archive material and irradiated material. Two complementary techniques were used, field emission gun scanning transmission electron microscopy (FEGSTEM) and Auger electron spectroscopy (AES). The unirradiated materials showed nonequilibrium segregation of chromium, molybdenum and phosphorus. Irradiation to intermediate fluences (~1.1021 n/cm2, E>l MeV) resulted in further enhancements, of silicon and phosphorus, at the grain boundaries. In the CP materials, molybdenum and chromium remained enhanced at the grain boundaries with chromium depletion on either side of the grain boundary, while chromium was significantly depleted in the HP material. At higher fluences (~1.1022 n/cm2, E>l MeV) chromium was significantly depleted in the CP Type 304 SS material. Molybdenum seemed to retain the chromium level at the grain boundary during irradiation up to a certain fluence. For the HP Type 304 SS material, which contained virtually no molybdenum, the chromium level was lower than for the CP heats.
Irradiation assisted stress corrosion cracking (IASCC) is a materials degradation phenomenon which affects light water reactor internals exposed to fast neutron radiation. IASCC appears as intergranular cracks, but does not require the formation of grain boundary chromium carbides, as in intergranular stress corrosion cracking (I GSCC) of thermally sensitized material. Service failures in various BWR, and in some cases also PWR, components of stainless steel and nickel-base alloys have been attributed to L4SCC.
IASCC has been observed in austenitic stainless steels in constant extension rate tensile (CERT) 3 hove an apparent threshold fast neutron fluence of 5.1020 n/cm2 (E> 1 tests~ 2, as well as in-service , a MeV). During irradiation by high energy neutrons microchemical and microstructural changes, as well as compositional changes caused by transmutations, are induced in the material. Although the specific mechanisms controlling IASCC have yet to be identified, any of these effects, as well as interactions between them, may be important for sensitization of the material. Moreover, neutron and gamma radiation produce changes of the reactor water which makes the environment more aggressive, i. e. radiolysis. Radiation induced segregation (RIS), which causes redistribution of alloying and impurity elements at the grain boundaries, has been suggested as an important factor for IASCC. Under neutron irradiation phosphorus, silicon and nickel segregate to the grain boundaries, while chromium and molybdenum become depleted. It seems probable that chromium depletion of grain boundaries due to RIS is the dominating mechanism, as BWR hydrogen water chemistry (HWC) has been shown to mitigate IASCC, at least at fluences below 2 to 3.1021 n/cm2 (E>l MeV).l RIS of phospho