Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Abstract

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370C, 400C and 450C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

“…. However, the microstructural characterization conducted in this program by means of synchrotron X-ray did not find evidence of LRO in specimens aged to 60 year of service equivalents [12]. Nevertheless, as described previously, aging to a 80-year service equivalent is in progress.…”

“…In its first year, the Argonne program focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions [12]. In addition to the aged Alloy 690, the study also involved specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa.…”

“…Figure15presents X-ray Diffraction (XRD) spectra in the three weld heats aged at 370°C for almost 75,000h, equivalent to 60 years of service (condition 370-Y60). Aged model Ni-33Cr alloy with known LRO[12] was included for reference. For the model alloy, in addition to the FCC peaks that are present in all conditions, the aged conditions show peaks -highlighted with green -from Pt 2Mo-type domains.…”

Light Water Reactor Sustainability: Program Materials Research Pathway (FY22 Technical Program Plan)

“…For FY 2021, comparisons have only been obtained on alloy X-750, and thus far, no obvious difference has been observed in SCC initiation and growth behavior between the KOH and corresponding reference LiOH water chemistries as shown in Figure 15. The third research area [28] was the evaluation of the effects of thermal aging and irradiation on microstructure and crack growth response of alloy 690. The objective of this research was to understand the microstructural changes occurring in high-Cr, Ni-based alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance.…”

Light Water Reactor Sustainability Program Materials Research Pathway Technical Program Plan

“…For FY 2021, comparisons have only been obtained on alloy X-750, and thus far, no obvious difference has been observed in SCC initiation and growth behavior between the KOH and corresponding reference LiOH water chemistries as shown in Figure 15. The third research area [26] was the evaluation of the effects of thermal aging and irradiation on microstructure and crack growth response of alloy 690. The objective of this research was to understand the microstructural changes occurring in high-Cr, Ni-based alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance.…”