Modelling nuclear fuel behaviour and cladding viscoelastic response

Research output: ThesisDoctoral ThesisCollection of Articles

Researchers

  • Ville Tulkki

Research units

Abstract

In light water reactors the nuclear fuel is in the form of uranium dioxide pellets stacked inside a thin-walled tube made from Zirconium alloy. The fuel rods provide the first barriers to the release of radioactivity as the isotopes are contained within the fuel matrix and the cladding tubes. Fuel behaviour analysis investigates the state of the fuel at given boundary conditions and irradiation history. The scope of this thesis consists of two main themes. The first is the uncertainty and sensitivity in fuel behaviour modelling and the tools required for its propagation to the rest of the nuclear reactor calculation chain. The second is the analysis and modelling of cladding response to transient stresses. A nuclear reactor is a strongly coupled system. The neutronics depend on the fuel and coolant temperature, the fuel temperature on the neutronics and the heat flux to coolant and coolant thermal hydraulics on the amount of heat transferred from the fuel. Propagation of uncertainties through the nuclear reactor calculation chain is an international on-going effort, and the complex interactions in the fuel rods make them challenging to analyze. In this thesis uncertainty and sensitivity of fuel behaviour codes is investigated and the development of a fuel module suitable for propagation of the uncertainties is detailed. The creep response of a cladding tube to changing conditions is conventionally modelled using the strain hardening rule. The rule assumes accumulated strain to be invariant during changes in the conditions, and is relatively simple to utilize. However, the original experiments which are used to justify the use of the strain hardening rule show that it applies only to a restricted set of conditions. In this thesis a simple methodology for predicting fuel cladding macroscopic response to stresses and imposed strains is developed by taking anelastic behaviour into account. The model is shown to perform well in describing both creep and stress relaxation experiments.

Details

Original languageEnglish
QualificationDoctor's degree
Awarding Institution
Supervisors/Advisors
Publisher
  • VTT Technical Research Centre of Finland
Print ISBNs978-951-38-8346-1
Electronic ISBNs978-951-38-8347-8
Publication statusPublished - 2015
MoE publication typeG5 Doctoral dissertation (article)

    Research areas

  • nuclear fuel behaviour, modelling, uncertainty and sensitivity analysis, cladding creep, stress relaxation, viscoelasticity

ID: 18411575