Abstract
Zirconium alloys are being commonly used as a material of choice for nuclear fuel claddings in water cooled nuclear reactors for decades due to their good corrosion resistance and low neutron absorption. However, the increasing operation conditions of next generation nuclear reactors (Gen-V) in terms of higher temperatures, pressures and higher neutron flux requires evaluation of further Zr cladding usability. The embrittlement of Zr claddings due to hydrogen pickup from reactor coolant is one of the issues for its potential use in Gen-IV reactors. Nanoindentation is an effective tool for analysis of the change of mechanical properties of hydrogen enriched Zr claddings from localised material volume. Zirconium alloy Zr-1Nb (E110) with experimentally induced hydrides was analysed by the means of nanoindentation. Zirconium hydrides were formed in the material after exposure in high temperature water autoclave. The optimized methodology of surface preparation suitable for nanoindentation is described and the resulting surface quality is discussed. The nanoindentation measurements were performed as an array of 10x10 indents across areas with hydrides. Depth dependent hardness and reduced modulus values measured by nanoindentation were compared between the material with no hydrogen content, low hydrogen content (127 ppm H) and high hydrogen content (397 ppm H). Complementary microhardness measurements at HV 0.1 were performed on all materials for bulk material hardness comparison.
Highlights
The nuclear fuel cladding serves as the first containment barrier of the fuel pellets in water-cooled nuclear reactors and subsequently in spent fuel storage casks
Due to their good mechanical properties, corrosion resistance and low neutron absorption, zirconium alloys are being commercially used as a nuclear fuel cladding material for decades
The duration of chemo-mechanical polishing surface finishing step was found to be crucial for correct nanoindentation measurement on E110 samples
Summary
The nuclear fuel cladding serves as the first containment barrier of the fuel pellets in water-cooled nuclear reactors and subsequently in spent fuel storage casks Due to their good mechanical properties, corrosion resistance and low neutron absorption, zirconium alloys are being commercially used as a nuclear fuel cladding material for decades. Hydrides formed in zirconium claddings affect the mechanical properties of the material, making it prone to cracking due to the internal stresses caused by fuel swelling and the propagation of fission gasses during the nuclear fuel cycle. This process of cracking associated with hydrides in Zr alloys is called Delayed Hydride Cracking (DHC) [2]. The comparison of mechanical properties between claddings with various amount of hydrogen content is assessed
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