Role of nitrogen in the structure and properties of proton irradiated 12Cr1MoWV ferritic/martensitic steel for advanced nuclear reactor cores, The

Abstract

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor, currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. Nitrogen content has been hypothesized to have a significant effect on the irradiated properties of alloy HT9. In this work, three otherwise similar alloys of HT9 with varying N contents (10 ppm, 190 ppm, 440 ppm N) are thoroughly characterized prior to irradiation with 1.5 MeV protons to 1 dpa of dose at 300 ̊C. In the unirradiated condition, the presence of ultrafine, intralath vanadium carbonitride (V(C,N)) precipitates are revealed for the first time in the 190 ppm and 440 ppm N alloys via centered dark field transmission electron microscopy (TEM). Lower N content result in finer intralath precipitates, whereas higher N content results in larger, elongated disks or needles. In addition to the quantitative assessment of interlath and intralath V(C,N) by TEM, thermodynamic simulations with ThermoCalc, and, for the first time, time-of-flight secondary ion mass spectrometry (ToF-SIMS) are utilized as complementary techniques, providing a high- throughput method for assessing trends in precipitate volume fractions. ToF-SIMS, combined with internal friction measurements, also provides the relative amount of interstitial N present across the three alloys. Characterization of the alloys shows that N content has a profound effect on the irradiated defect structures. On-zone scanning TEM (STEM) is used to determine that, as N content increases, the average dislocation loop diameter decreases, while the number density of loops increases, a behavior consistent with a reduction in self-interstitial atom (SIA) cluster mobility. Additionally, STEM energy dispersive spectroscopy finds extensive Ni clustering on dislocations and V(C,N) interfaces. The Mid and High N specimens exhibit significantly less hardening relative to the Low N sample. The decrease in hardening is attributed to the presence of V(C,N), which provides an alternative short-range site beyond dislocation loops and line dislocations for the formation of Ni clusters, resulting in fewer Ni clusters on dislocations in the Mid and High N alloys. The data indicates increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the SIA cluster mobility in F/M steels for different temperature and dose applications.