Modeling hydrogen localization in Zircaloy cladding subjected to temperature gradients

Katheren R.B. Nantes, Miaomiao Jin, Arthur T. Motta

Research output: Contribution to journalArticlepeer-review


In light water reactors, Zr-based alloys used for nuclear fuel cladding undergo oxidation and absorb hydrogen, which can precipitate as brittle zirconium hydrides. During reactor normal operation conditions, hydrogen concentration varies locally due to various factors, where local high concentration can significantly degrade cladding mechanical properties and, hence, its service life. This study aims to quantify the redistribution of hydrogen within the fuel rods caused by temperature gradients, considering geometric irregularities such as inter-pellet regions, oxide spallation, and missing pellet surfaces. Through simulations, we have discovered that these temperature variations can lead to significant local hydrogen enrichment, even under normal operational conditions. Consequently, the Zircaloy cladding may suffer from critical weakening in its mechanical performance due to excessive hydride precipitation at specific locations. These findings underscore the importance of accounting for local hydrogen concentrations when evaluating the overall reliability of the cladding.

Original languageEnglish (US)
Article number154853
JournalJournal of Nuclear Materials
StatePublished - Feb 2024

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • General Materials Science
  • Nuclear Energy and Engineering

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