Abstract
Plasma nitridation was conducted to modify the surfaces of Zircaloy-4. Scanning electron microscopy (SEM), transmission electron microscopy (TEM), and Raman analysis were used to characterize microstructures and phases. Surface indentation and cross-sectional indentation were performed to evaluate mechanical property changes. Nitridation forms a thin layer of ZrN phase, followed by a much deeper layer affected by nitrogen diffusion. The ZrN phase is confirmed by both TEM and Raman characterization. The Raman peaks of ZrN phase show a temperature dependence. The intensity increases with increasing nitridation temperatures, reaches a maximum at 700 °C, and then decreases at higher temperatures. The ZrN layer appears as continuous small columnar grains. The surface polycrystalline ZrN phase is harder than the bulk by a factor of ~8, and the nitrogen diffusion layer is harder by a factor of ~2–5. The activation energy of nitrogen diffusion was measured to be 2.88 eV. The thickness of the nitrogen-hardened layer is controllable by changing the nitridation temperature and duration.
Highlights
Zirconium-based alloys are widely used as fuel cladding and in-core structural components in light water reactors due to their low neutron absorption cross-section, good corrosion resistance, and high strength [1]
The advantages of Zr fuel cladding are countered by its anisotropic properties and poor oxidation response in steam at high temperatures
11 study compares both diffusivities the previous obtained from the N ion layer by changing nitrogen diffusion by adjusting nitridation temperature and time
Summary
Zirconium-based alloys are widely used as fuel cladding and in-core structural components in light water reactors due to their low neutron absorption cross-section, good corrosion resistance, and high strength [1]. While Zr cladding exhibits high corrosion resistance under normal operating conditions, its vulnerability to hydrogen pickups and rapid oxidation are well-known issues in the event of a loss of coolant accident (LOCA). Reactions between the Zircaloy fuel cladding and water vapor result in abnormal oxidation, and subsequent hydrogen production, as occurred in the Fukushima Daiichi nuclear accidents. One technique envisioned to avoid the above scenarios is to coat fuel cladding tubes with an oxidation-resistant and thermally stable material [2]. An ideal coating is expected to reduce oxidation rates under abnormal conditions to avoid possible hydrogen explosions when Zircaloy-4 is exposed to super-hot steam. The easy swelling of Cr at relatively low damage levels [6], each amorphization of MAX phase alloys (Ti2 AlC as one example) at interfaces [7], and complicated interaction reactions between
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