Development and evaluation of fuel performance analysis code FuSPAC

• Published in: International Journal of Advanced Nuclear Reactor Design and Technology Vol. 4; no. 3; pp. 129 - 138
• Main Authors: Cong, Tenglong, Chen, Hao, Chen, Chao, Zhang, Shaohong, Gu, Hanyang
• Format: Journal Article
• Language: English
• Published: KeAi Communications Co., Ltd 01-09-2022
• Subjects: Fuel behaviour under irradiation; Fuel performance analysis code; FuSPAC; UO2 fuel
• ISSN: 2468-6050; 2468-6050
• DOI: 10.1016/j.jandt.2022.08.002

The nuclear fuel rod should maintain self-sustainable nuclear fission, as the first safety barrier to prevent the leakage of radioactive products. The fuel rod performance analyses are essential to ensure reliable long-term operation of fuel rod. In this work, the fuel performance analysis code FuSPAC is developed based on the commonly used 1.5-dimensional method for UO2 ceramic fuel with Zirconium cladding. The physical and numerical models used in FuSPAC code are described. By comparing the predicted values from FuSPAC code with the available in-reactor fuel rod experimental data and the numerical results from the well-recognized FRAPCON code, the capacity of FuSPAC code on the prediction of fuel pin and cladding under long term irradiation conditions are validated from five aspects, including fission gas release behaviour, thermal behaviour, rod internal void volume, cladding corrosion and cladding hoop strain, showing good accuracy in the behaviour analysis of the irradiated UO2 fuel rod.