Abstract
Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.
Highlights
Accepted: 26 January 2021With sources of fossil fuels becoming increasingly exhausted and the environmental challenges associated with their use, efficient and clean nuclear energy will become a promising energy source in the future
The temperature field distribution in °C of the fuel elements after irradiation for 1200 days is shown in Figure 2, which are pure Zr, N36, Zr-4, and Zr-2.5Nb-0.5Cu
Mises stress field distributions in the MPa of cladding for different alloy systems during stable operation for 1200 days under irradiation conditions; these data can be employed to evaluate the cladding safety. It may be observed from the simulation results that the stress field distribution of the cladding was similar to the temperature field distribution; namely, the stress gradually decreased from the inner to the outer surface of the cladding in the radial direction
Summary
With sources of fossil fuels becoming increasingly exhausted and the environmental challenges associated with their use, efficient and clean nuclear energy will become a promising energy source in the future. Cladding is one of the most important components in a nuclear reactor It has been used in extreme environments with high radiation levels, high temperatures, and high pressure levels for a long time [1]. Zirconium (Zr) alloys have been applied as a cladding material because of their high temperature mechanical properties, small neutron absorption cross section, strong radiation resistance, and good irradiation stability [2]. The effects that extreme environments—e.g., with high irradiation levels, high temperatures, and high pressure levels—have on the mechanical and thermal properties of Zr alloy cannot be determined through conventional experimental methods [5,6]. The user subroutine UMAT of ABAQUS finite element software was employed to study the radiation-thermal-mechanical coupling behavior of fuel rods by Tang et al [9]. The effect of neutron irradiation on the mechanical and thermal properties of Zr and Zr alloy, as well as the temperature field, displacement field, and stress field of Zr alloy cladding under neutron irradiation, were studied using the ABAQUS distribution and evolution mechanism
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