Hui Zhao
Current Organizations / Institutes
The organizations/institutes from the scholar's latest article
Research Subjects
Hui Zhao has a total of 7 co-authors. Here are the top co-authors by number of shared publications:
- Baodong Shi
State Key Laboratory of Crane Technology, Yanshan University, Qinhuangdao, 066004, China
1 shared publication - Chong Yang
National Engineering Research Center for Equipment and Technology of Cold Rolled Strip, School of Mechanical Engineering, Yanshan University, Qinhuangdao 066004, China
1 shared publication - Dongxu Guo
National Engineering Research Center for Equipment and Technology of Cold Rolled Strip, School of Mechanical Engineering, Yanshan University, Qinhuangdao 066004, China
1 shared publication - Jianjun Mao
The First Institute, Nuclear Power Institute of China, Chengdu 610041, China
1 shared publication - Jiantao Qin
The First Sub-Institue, Nuclear Power Institute of China, Chengdu 610005, China
1 shared publication - Lu Wu
State Key Laboratory of Nuclear Fuel and Materials, The First Sub-Institute, Nuclear Power Institute of China, Chengdu 610041, China
1 shared publication
Latest Publications
- Open Access
- 1 February 2021
- Published by MDPI AG in Applied Sciences
- Vol. 11 (3) , 1308
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 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.
- 1 February 2021
- Published by MDPI AG in Applied Sciences
- Vol. 11 (3) , 1308
Open AccessCoupled 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 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.