Numerical study on cladding burst behavior considering irradiation and oxidation with the multiphysics coupling method
This study investigates alterations in the performance of fuel rod cladding during loss of coolant accident through the development of a two-dimensional axisymmetric burst analytical model using COMSOL Multiphysics. The model incorporates irradiation and oxidation effects while accounting for cladding fluid-solid heat transfer, high-temperature anisotropic creep with gap conductivity, and temperature-time dependent oxidation and irradiation processes. This comprehensive approach enables multi-physics field coupling of heat-fluid-solid-irradiation-oxidation interactions. To elucidate the complex oxidation effects, the analysis separates three key components: oxidative oxygen uptake, crack depth, and degree of hydrogenation, examining their individual impacts on cladding behavior. The results indicated that under irradiation conditions with fast neutron flux of 2 × 1018∼8 × 1018 n/m2s, the model predicts an acceleration of burst time by approximately 8 s and a 41 % reduction in burst strain, attributed to lattice structure deterioration from irradiation-induced defects. The oxidation analysis shows that when the oxygen uptake increases from 3 % to 20 %, the burst time is delayed. This clearly demonstrates the protective role of the ZrO2 layer. Crack depth analysis shows that increasing depth from 0.1 mm to 0.5 mm substantially reduces both cladding burst strain and strength. Hydrogen content shows a phase-dependent effect: in the low-temperature α-phase, it decreases cladding plastic toughness, with 600 ppm hydrogen reducing the burst strain to about 10 % at 900 K and 1000 ppm hydrogen reducing the burst strain to 5 %, while enhancing performance in the high-temperature β-phase. These findings provide crucial theoretical support and reference data for nuclear reactor safety design and operation, contributing significantly to the understanding of cladding behavior under complex operating conditions.
Authors
- Junjie Yu
- Shangke Li
- Runhua Gao
- Gen Li
- Xin Zhang
Publication Details
- Journal
- International Journal of Advanced Nuclear Reactor Design and Technology
- Published
- 2026-01-14
- DOI
- https://doi.org/10.1016/j.jandt.2026.01.003
- Primary Topic
- Nuclear Materials and Properties
- Type
- article
- Field-Weighted Citation Impact
- 0.00