Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

Abstract

The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10 19 n/cm 2 , 2.0 x 10 20 n/cm 2 , and 1.0 x 10 21 n/cm 2. These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10 19 n/cm 2 , comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60-80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project. Following completion of the irradiations at ATR in the water-cooled center flux trap, irradiation assisted stress corrosion cracking (IASCC), fracture toughness (FT), tensile testing, and transmission electron microscopy (TEM) studies were conducted at INL. Testing under normal water chemistry (NWC) conditions indicated a negligible effect on crack growth rate (CGR) at the medium and high fluence irradiation levels compared to the CGRs measured in unirradiated material. Under hydrogen water chemistry (HWC) conditions, only a modest increase in CGR was measured at the high fluence level compared to the medium fluence level. Tensile and FT measurements were far more sensitive to irradiation level, especially in the case of alloy X-750, with an increase of 21, 29, and 57% of yield strength compared to unirradiated levels for the low, medium, and high fluences, respectively, and an approximate 4, 18, and 41% reduction in FT for the low, medium, and high fluences, respectively. Additionally, TEM analyses showed an increase in dislocation loop size as a function of irradiation dose for both alloys, but a negligible change in loop density.