Short communicationThe effect of cooling rate and grain size on hydride microstructure in Zircaloy-4
Section snippets
Main body
Zirconium alloys are used in the nuclear industry as fuel cladding, as it has a good strength to neutron absorption cross section ratio and reasonable corrosion resistance. One concern when using zirconium alloys in high temperature water reactors is corrosion, as during service it can react with high temperature water to generate an oxide scale and pick up hydrogen [1]. For service conditions (∼350 °C) this hydrogen may exist in solution where it is highly mobile [2,3]. The hydrogen travels
Summary
We observe that the structure and population of hydrides is strongly determined by the cooling rate when hydrides are precipitated in zirconium. The relative grain size, and thus grain boundary area vs grain interior, can significantly change the hydride population. Importantly in terms of understanding the performance of nuclear fuel, the population of hydrides may influence the failure of the zirconium cladding during delayed hydride cracking and/or long-term storage.
Acknowledgements
TBB acknowledges funding from the Royal Academy of Engineering for his research fellowship. TBB and VT acknowledge funding from EPSRC through the HexMat programme grant (EP/K034332/1). Electron microscopy was performed within the Harvey Flower Electron Microscopy Suite and the Quanta was purchased within the Shell AIMS UTC. We would like to thank Alex Foden for assistance with the EBSD pattern matching.
References (19)
- et al.
External corrosion of cladding in Pwrs
Nucl. Eng. Des.
(1975) - et al.
High-temperature oxidation and quench behaviour of Zircaloy-4 and E110 cladding alloys
Prog. Nucl. Energy
(2010) - et al.
Zirconium hydride precipitation kinetics in Zircaloy-4 observed with synchrotron X-ray diffraction
J. Nucl. Mater.
(2015) - et al.
EBSP measurements of hydrogenated Zircaloy-2 claddings with stress-relieved and recrystallized annealing conditions
J. Nucl. Mater.
(2006) - et al.
Hydride formation on deformation twin in zirconium alloy
J. Nucl. Mater.
(2016) - et al.
Hydride reorientation in Zircaloy-4 cladding
J. Nucl. Mater.
(2008) - et al.
Effect of thermo-mechanical cycling on zirconium hydride reorientation studied in situ with synchrotron X-ray diffraction
J. Nucl. Mater.
(2013) - et al.
Hydrogen in zircaloy: mechanism and its impacts
Int. J. Hydrogen Energy
(2015) - et al.
Atomic scale analysis of grain boundary deuteride growth front in Zircaloy-4
Scripta Mater.
(2018)
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