NANOINDENTATION OF HYDROGEN ENRICHED Zr-1Nb ZIRCONIUM ALLOY NUCLEAR FUEL CLADDINGS

Ondřej Libera, Patricie Halodová, Petra Gávelová, Jakub Krejčí

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.

Keywords


Nanoindentation, nuclear fuel claddings, Zr-1Nb

Refbacks

  • There are currently no refbacks.


Creative Commons License
This work is licensed under a Creative Commons Attribution 4.0 International License.

ISSN 2336-5382 (Online)
Published by the Czech Technical University in Prague