Evaluation of Nuclear Fuel Centerline Temperature Using New UO2 Thermal Conductivity Models

Abstract

Abstract:The nuclear industry needs of prediction of behavior and life-time, for a wide range of normal, off-normal and accident conditions for safe and economic operation. Among different thermo-mechanical properties that can be predictable, the knowledge on the radial temperature distribution of the UO 2 (uranium dioxide) nuclear fuel during the operation of nuclear reactors is essential for safety as different mechanical and thermal-hydraulic thresholds should be respected. One of the attributes of the Brazilian CNEN (Nuclear Energy Commission) is to assess the performance of the fuel rods used in these reactors in high-burnup regimes. The effective removal of the heat generated in the fuel rods constitutes one of the primary points to consider in the design of nuclear reactors. One of the important physical parameters in the study of heat conduction from the nuclear fuel to the coolant in a PWR (pressurized water reactor) is its thermal conductivity. It is therefore desirable that the empirical models, updated for the calculation of thermal conductivity in the fuel region be developed from new sets of experimental data from the irradiated fuel rods in controlled environments. This paper presents the obtained results of implementing of a new model for thermal conductivity of the UO 2 in the FRAPCON code.