B12-P-09Characterization of zirconium hydrides in Zircaloy-4 cladding with respect to cooling rate

Abstract

Zircaloy-4 has been of great interest as the core material for nuclear fuel cladding and fuel assembly in the pressurized water reactor (PWR) since it does not only have substantially low neutron absorption during the nuclear power generation, but also show superior mechanical properties including high corrosion resistance in the radiation environment [ 1 ]. However, excessive hydrogen (H) is precipitated as the zirconium (Zr) hydride with the form of platelets in the Zr cladding when the H concentration exceeds the terminal solid solubility (TSS), which brings about reduction of ductility and fracture toughness, and subcritical crack growth related with delayed hydride cracking (DHC). Therefore, it is essential to characterize the Zr hydrides in terms of the morphology and orientation under the applied stress and temperature conditions. In this study, three different Zr hydrides including δ-hydride, γ-hydride, and surface hydride are produced by controlling the cooling rate from 0.5 to 300°C/min to characterize habit planes and crystallographic relation between δ -matrix and Zr hydrides in the Zircaloy-4 cladding. TEM investigations have been conducted to analyze detailed crystallographic relation in which changes in the selected area electron diffraction (SAED) patterns are examined with a TEM heating holder during the sequential heating and cooling process with the same TEM specimen.