Hydrogen and Oxygen Distribution During Corrosion of Zirconium Alloy Nuclear Fuel Cladding

  • Christopher Jones

Student thesis: Phd

Abstract

Zirconium alloys are widely used in light water reactors as fuel cladding, acting as the primary barrier between the nuclear fuel and the coolant, due to the excellent neutron transparency, acceptable corrosion resistance and good mechanical properties of the zirconium. While in service these alloys degrade due to exposure to high temperature aqueous environments and strong radiation fields. As these Zr alloy components often perform critical safety roles, their degradation in service must be understood. In this thesis I use the capabilities of the NanoSIMS 50L, a magnetic sector mass spectrometer capable of nanoscale lateral resolution, to analyse the distribution of hydrogen and oxygen in corroded Zircaloy-4 and Zircaloy-2 and the impact that irradiation has on oxide growth and distribution of oxygen in the oxide layer. In this study an array of H+ irradiated samples was produced. Two batches of these samples were oxidised in simulated pressurised water reactor coolant for 52 and 131 days before being irradiated at the Dalton Cumbrian facility at 350 °C with 1 MeV H+ ions using the BABY beamline to doses of 0.25 and 0.75 dpa. These samples were then returned to an autoclave isotopically spiked with 5% H2 18O and 50% 2H2O at 320 °C for forty days. This produced an array of samples with oxides that were pre-transition (92 days) and at the point of transition (171 days). Alongside these samples were unirradiated samples, isotopically spiked with 2H, that were made available by Jacobs plc. Samples were analysed with NanoSIMS and correlative techniques and the results of these analyses are presented in three papers. In the first of these papers I show that oxygen transport across oxide layers is dominated by transport along cracks and pores, while hydrogen diffusion appears to be lattice diffusion based and limited in extent. Additionally, I apply a technique to image areas sequentially with the NanoSIMS that allows for analysis of more material by an order of magnitude than typical analysis. In the second paper I show that hydrogen segregates to the interfaces between secondary phase particles, most likely Zr(Fe,Cr)2 intermetallics, in the zirconium base metal in Zircaloy- 2 and Zircaloy-4. Additionally, I show that prolonged analysis with NanoSIMS leads to the accumulation of subsurface irradiation damage in the sample which destroys hydrogen trapping sites. 15 | P a g e In the final paper, which focusses on the impact of irradiation on oxygen diffusion and zirconium oxide growth, I show that the impact of proton irradiation varies greatly depending on the oxide thickness during irradiation. For pre-transition oxides the oxide growth rate and oxygen diffusivity are increased, but this leads to inhomogeneous oxide growth. When oxides approaching transition are irradiated this leads to increased cracking and spallation of the oxide layer.
Date of Award31 Aug 2021
Original languageEnglish
Awarding Institution
  • The University of Manchester
SupervisorMichael Preuss (Supervisor) & Katie Moore (Supervisor)

Keywords

  • Oxygen
  • Hydrogen
  • Irradiation
  • Nuclear Fuel Cladding
  • Zirconium Oxide
  • Zirconium
  • NanoSIMS

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