The embrittlement of reactor pressure vessels (RPV) in light water reactors (LWR) during service remains a significant safety concern. Predicting mechanical properties changes during service relies on accurate microstructural information. However, most microstructural information and understanding on the effects of irradiation on microstructure has focused on nanoscale Cu and MNS precipitation within grain interiors. A lot less is known about the evolution of dislocations and grain boundaries (GBs) during irradiation, despite their roles in hardening and embrittlement. Therefore, this project addresses chemistry and character evolution of grain boundaries in RPV steels. This project will more specifically focus on a Cu-free, high Ni, high phosphorous RPV steel that was irradiated as part of the ATR UCSB ATR-2 experiment. This steel is part of a larger alloy series, whose irradiated microstructures were initially characterized as part of previous post irradiation experiment (PIE). The data showed a surprisingly high fraction of low angle grain boundaries (LAGBs) captured within the atom probe tomography (APT) datasets. The low angle character was inferred from the tight arrays of solute (Mn, Ni, Si, P) decorated dislocations present in the grain boundary planes. Moreover, contrary to expectations, comparable segregation was measured from the presumed LAGBs and the presumed high angle grain boundaries (HAGB). However, no information on the microstructure prior to irradiation or on the grain boundary characters after irradiation had been collected, severely limiting the interpretation of the impact of irradiation on the RPV microstructure. Therefore, we propose to compare grain boundary character and chemistry before and after irradiation to quantify the change in segregation induced by irradiation and correlate chemistry with grain boundary character. While we cannot characterize the same GB before and after irradiation, a statistical analyses of grain boundary distributions will be obtained using high resolution energy backscattered diffraction (EBSD) imaging and analysis to determine if irradiation induces changes in GB character distribution. Furthermore, using EBSD to select specific grain boundaries, we will quantify potential solute segregation and precipitation at unirradiated GBs and at comparable irradiated GBs using APT. We will focus on several LAGBs of known misorientation to address possible dependence with grain boundary plane as well as several HAGB to compare segregation level and propensity for GB precipitation. From APT analyses performed in a previous RTE, the selected steel exhibits a high density of precipitates in grain interiors as well as precipitates on isolated dislocations, offering the additional opportunity to quantify differences in the irradiation response, specifically segregation and precipitation at isolated dislocation in grain interiors versus GB dislocations. The microscopy work will be performed at the Center for Advanced Energy Studies (CAES) and the anticipated period for this award is six months.

The embrittlement of reactor pressure vessels (RPV) in light water reactors (LWR) during service remains a significant safety concern, particularly in the context of life extension programs. Upon irradiation, microstructural changes that include nanoscale precipitation of Cu-rich and Ni, Mn, and Si (MNS) rich phases contribute to significant hardening, increase in the brittle to ductile transition temperature, and decrease in fracture toughness. Predicting mechanical properties changes during service relies on accurate microstructural information. Several DOE-NE programs have addressed and continue to build predictive embrittlement models and a quantitative understanding of radiation damage. However, most microstructural information and understanding on the effects of irradiation on microstructure has focused on nanoscale Cu and MNS precipitation within grain interiors. A lot less is known about the evolution of dislocations and grain boundaries during irradiation, despite their roles in hardening and embrittlement. Moreo-ver, recent work revealed Cu-MNS precipitation at dislocations and grain boundaries (GB), highlighting significant synergy between defects. Therefore, this project will address this knowledge gap and focus on quantifying chemistry and character evolution of grain boundaries in an RPV steel. The results will contribute to improving microstructural databases used to inform predictive models of embrittlement that contribute to the continued operation of current U.S. nuclear reactors. This proposal builds on several National Scientific User Facility (NSUF) awards. The selected material was irradiated as part of an irradiation experiment performed as part of the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) NSUF. The experiment was awarded to University of California, Santa Barbara (UCSB) several years ago with full funding for the irradiation experiment in the ATR provided by DOE through the NSUF. This proposal also builds on the microstructural data collected during a previous RTE awarded in 2017 and that initiated the characterization of a series of RPV steels containing various concentrations of Cu, Ni and P.