Environmentally Assisted Cracking of LWR Materials * 0

Research on environmentally assisted cracking (EAC) of light water reactor materials has focused on (a) fatigue initiation in pressure vessel and piping steels, (b) crack growth in cast duplex and austenitic stainless steels (SSs), (c) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (d) EAC in high-nickel alloys. The effect of strain rate during different portions of the loading cycle on fatigue life of carbon and low-alloy steels in 289OC water was determined. Crack growth studies on wrought and cast SSs have been completed. The effect of dissolved-oxygen concentration in high-purity water on IASCC of irradiated Type 304 SS was investigated and trace elements in the steel that increase susceptibility to intergranular cracking were identified. Preliminary results were obtained on crack growth rates of high-nickel alloys in water that contains a wide range of dissolved oxygen and hydrogen concentrations at 289 and 32OOC. The program on Environmentally Assisted Cracking of Light Water Reactor Materials is currently focused on four tasks: fatigue initiation in pressure vessel and piping steels, fatigue and environmentally assisted crack growth in cast duplex and austenitic SS, irradiation-assisted stress corrosion cracking of austenitic SSs, and environmentally assisted crack growth in high-nickel alloys. Measurements of corrosion-fatigue crack growth rates (CGRs) of wrought and cast stainless steels has been essentially completed. Recent progress in these areas is outlined in the following sections. Irradiation Assisted Stress Corrosion Cracking Failures of reactor-core internal components in both BWRs and PWRs have occurred after accumulation of relatively high fluence (>5 x 1020 ncm-2, E >1 MeV). As neutron fluence increases, initially nonsensitized austenitic SSs can become susceptible to intergranular failure. This type of degradation is commonly referred to as irradiation-assisted stress corrosion cracking (IASCC). Although most failed components can be replaced, some components would be very difficult or impractical to replace. Heat-to-heat variations in susceptibility to IASCC have been observed to be very significant, even among high-purity (HP) materials containing virtually identical chemical compositions. Although radiation-induced grain-boundary Cr depletion is believed by most investigators to play an important role in IASCC, additional deleterious processes may be associated with trace impurities that are not usually identified in material specifications. Such trace elements could be introduced during steelmaking processes or during fabrication and welding. * Job Code A2212; NRC Program Manager: Dr. M. McNeil A Kodsma etal.. CP 304 dry lube, 1 3 x 10' n cni this study. CP 304-A absorber tube. 2.0 x IO' n cm' this study. HP 304 ebswber tube, 1.4 x lo2 ' n cm' ' Figure 1. . Percent IGSCC vs. DO for H P and CP Type 304 SS neutron-absorber tubes (this study) and CP Type 304 SS dry tubes (Kodama et al.). Three distinct trends in DO dependence are evident. Dissolved Oxygen (ppm) Slow-strain-rate-tensile (SSRT) tests in simulated B W R environments on highand commercialpurity (HP and CP) Type 304 SS were performed to determine the effects of water chemistry on susceptibility to IASCC. To obtain some insight into the mechanism(s) of IASCC, we attempted to correlate the susceptibility to intergranular cracking determined by SSRT tests with results of microchemical analysis of grain boundaries by Auger electron spectroscopy (AES). The effect of dissolved oxygen (DO) in HP water on the susceptibility of irradiated materials to IASCC is shown in Fig. 1 for HP and CP neutron-absorber tubes and a control-blade sheath (fabricated from another CP-grade heat). Negligible IGSCC was observed in specimens from the control-blade sheath for all fluence levels; therefore, no tests were conducted to investigate the effect of DO, other than at ~0.3 ppm. Results obtained by Kodama et al.,l*2 are also shown for comparison. The effect of electrochemical potential (ECP) on the susceptibility of irradiated materials to IASCC is shown in Fig. 2 along with some results by Indig et al.3 The effect of DO level and ECP on IASCC appears to be different for the HP and CP materials. The HP heats were less sensitive to DO level and ECP and were more susceptible to IASCC than the CP heats for all DO and fluence levels. No IASCC was observed in the CP heats for ECP <-140 mV SHE and DO <0.01 ppm. The HP absorber specimen that exhibited a surprisingly high susceptibility to IGSCC of ~ 1 8 % at a very low DO of -0.002 ppm (Fig. 1) and ECP of -320 mV SHE (Fig. 2) was examined in detail by AES. An intergranular (IG) region on the fracture surface was sputtered to a depth of -60 nm, and AES spectra were obtained as a function of depth from the fracture surface. Unexpected concentrations of F, Ca, B, Zn, and A1 were observed in the AES spectra. A1 was used as a deoxidizer during melting of the HP-grade SS and Ca could have originated from CaF2 that may have been used as flux in melting HP SS. The B and Zn on the fracture surface was probably picked up from the water. As shown in Fig. 3, a higher level of fluorine on grain boundaries may be associated with higher susceptibility to IASCCP Inadvertent contamination of reactor components by fluorine could occur during pickling (in a solution containing HF) in the case of tubular components such as neutron-absorber-rod tubes, and by an F-containing weld flux in the case of large welded components such as BWR core shrouds and certain older top guides. A synergistic effect of a lower concentration of Cr and a higher concentration of fluorine on grain boundaries on susceptibility to IGSCC is consistent with IGSCC results by Ward et al.5 Figure 2. . Percent IGSCC vs. ECP of H P and CP Type 304 SS neutron-absorber tubes (this study) and CP Type 304 SS BWR sheet material (Indig et al.)