Development and Verification of Performance Analysis Code for Fuel Element of Sodium-cooled Fast Reactor

For many years, sodium-cooled fast reactors have occupied the most important part of the closed fuel cycle. In order to improve the economy of sodium-cooled fast reactors, the nuclear industry around the world is actively increasing fuel burnup as much as possible. The behavior simulation of fuel elements under high fuel burnup is a key issue in the design and reliability of fuel elements. In this case, it is necessary to develop computer code that can accurately analyze fuel behavior to evaluate the behavior and reliability of high-fuel fuels, and as a safety analysis tool to evaluate the performance and behavioral evolution of fuel elements under steady-state, transient and accident conditions. For the above reasons, the Chinese Institute of Atomic Energy has developed FIBER, a performance analysis code for fuel elements of sodium-cooled fast reactor. The code consists of two main parts: The first part is used to analyze the temperature distribution, the thermal deformation and fission gas release; The other part is used to analyze the mechanical behavior of fuel elements. In the thermal analysis part, the axisymmetric finite volume method is applied to the entire length of the fuel element. The code has the ability to calculate thermal conductivity, gap heat transfer, coolant heat transfer, fission gas release, fuel restructure, solid fission product migration, and plenum pressure. In the mechanical analysis part, the axisymmetric finite element method is applied to the entire length of the fuel elements. The code can simulate the phenomena of thermal expansion, densification, irradiation swelling, pellet cracking, elasticity ,plasticity, creep, and PCMI. The thermal analysis part and the mechanical analysis part are coupled, and the convergence of temperature and deformation is obtained in each time step through iteration. FIBER code consists of many theoretical models, empirical models, and parameters that control the calculation process. However, fuel behavior cannot be explained only by a simple combination of these models, because fuel behavior is the result of the coupling of many phenomena. Therefore, as many cases as possible must be used for code verification to determine the appropriate model and parameter selection. The irradiation data of UO2 and MOX of the Russian BN600 reactor were obtained through research. The two fuel elements operated in the Russian BN600 for 559 days, with maximum fuel burnup of 11.8at% and maximum irradiation damage of 78 dpa. The FIBER code was used to analyze the above two fuel elements, the calculation results of fission gas release rate, irradiation deformation, gap, columnar region, are compared with the irradiation data. The comparison results show that the FIBER code is effective for evaluating the irradiation deformation, columnar crystal region, and fission gas release performance of high burnup fuel elements. © 2024 Atomic Energy Press. All rights reserved.