Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it service for long time under the effects of neutron radiation with high temperature and pressure, which results thermomechanical coupling behavior during process. Focusing on UO2/Zr elements, macroscopic responses pure Zr, Zr-Sn, Zr-Nb binary system alloys, as well Zr-Sn-Nb ternary materials, were studied irradiation. As heat source, thermal conductivity expansion coefficient models UO2 core established, an irradiation growth model Zr multisystem was built. Based user subroutine (UMAT) ABAQUS, current theoretical implemented into finite element framework, consequent calculated. The distribution temperature, stress field cladding, their evolution over analyzed. It found that displacement sensitive to alloying elements due irradiated growth.