Thermochemical and thermomechanical modeling of nuclear fuel: Dissertation

Henri Loukusa

Research output: ThesisDissertationCollection of Articles


Nuclear fuel undergoes various changes in its properties and composition as it is irradiated in a nuclear power plant. Typical light-water reactor fuel consists of ceramic uranium oxide pellets encased in a zirconium alloy tube called the cladding. As actinide atoms are fissioned, the elemental composition of the pellets change as fission products form. Irradiation damage is accumulated and the pellets expand due to the accumulation of fission products. The cladding also experiences irradiation damage, and undergoes mechanical deformation during irradiation due to a pressure differential between the inside of the rod and the coolant and mechanical interaction with the expanding pellet.
The effects of the composition change on the chemical properties are studied in this thesis with thermochemical modeling using Gibbs energy minimization. Especially the oxygen potential of the pellets is studied, as it is a general indicator of the oxidation state and relatively easily measurable experimentally. Validation of the oxygen potential predictions is performed against experimental data from the literature for fresh and irradiated fuel. In addition, the release of corrosive gases from the pellet is studied, as they may affect cladding crack formation.
The FINIX fuel behavior module has been developed at VTT for multiphysics applications, and in this thesis its thermomechanical modeling capabilities are extended to cover long irradiation periods. Several models of important phenomena that occur during long irradiations were not present in FINIX, but were implemented as part of this thesis in order to improve the temperature predictions of FINIX. As a result of this development, the temperature predictions during long irradiations have improved substantially in accuracy.
Additionally, uncertainty analysis based on the order statistics method was performed on fission gas release predictions of the ENIGMA fuel performance code. Fission gas release can be correlated with the instant release fraction of some radionuclides in spent fuel disposal. The instant release fraction is the fraction of radionuclides that are released relatively instantly as groundwater becomes in contact with the fuel after the disposal canister has failed. With an estimate of the uncertainty of the fission gas release predictions, the uncertainty of the instant release fraction can be studied in more detail.
Original languageEnglish
QualificationDoctor Degree
Awarding Institution
  • Aalto University
  • Tuomisto, Filip, Supervisor, External person
  • Tulkki, Ville, Advisor
Award date25 Feb 2020
Print ISBNs978-952-60-8942-3
Electronic ISBNs978-952-60-8943-0
Publication statusPublished - 25 Feb 2020
MoE publication typeG5 Doctoral dissertation (article)


  • nuclear fuel behavior
  • thermochemistry
  • thermomechanics
  • uncertainty analysis


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