Study of Severely Hydrided Zr-4 PHWR Fuel Clad Using Ring Tensile Test

Abstract

Zircaloy (earlier Zr-2 and now Zr-4) is widely used as cladding alloy for nuclear fuel elements in pressurised heavy water reactor. High thermal stresses, neutron irradiation and corrosion in the reactor change the material properties of the clad. One such important factor for changing the mechanical properties of clad is hydrogen pickup by clad tubes. The stress imposed on the cladding is mostly hoop stress on the circumference of the tube and therefore the data on the circumferential strength and ductility of the nuclear fuel cladding are required. Hydrogen content has an effect on mechanical properties of clad. In this study an attempt has been made to study the effect hydrogen content (~45-50 ppm in as-fabricated and hydride clads had around 392, 780, and 1200 wppm hydrogen content in homogenised samples and hydrogen contents 393 wppm and having hydride layer thickness 11.8 microns and 480 wppm and having hydride layer thickness 16.2 microns for hydride rim samples) on the transverse mechanical properties of the Indian PHWR Zircaloy clads. It was observed that there was no change in mechanical properties for the clad tubes with rimmed hydrides. In case of uniformly distributed hydrides, upto 400 wppm hydrogen content, there is not much change in elongation. Thereafter, break strain decreases as we increase the hydrogen content of the uniformly hydrided clad tubes.