Journal of Energy and Power Engineering 8 (2014) 1054-1058 Evaluation of Nuclear Fuel Centerline Temperature Using New UO 2 Thermal Conductivity Models Daniel Artur Pinheiro Palma 1 , Amir Zacarias Mesquita 2 , Franciole da Cunha Marinho 3 and Marcelo da Silva Rocha 4 1. CNEN (Brazilian, Nuclear Energy Commission), Rio de Janeiro 22290-901, Brazil 2. CDTN (Nuclear Technology Development Center), Belo Horizonte 31270-901, Brazil 3. UFRJ (Federal University of Rio de Janeiro), Macaé 27930-560, Brazil 4. IPEN (Nuclear and Energy Research Institute), São Paulo 05508-000, Brazil Received: December 03, 2013 / Accepted: January 10, 2014 / Published: June 30, 2014. 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. Key words: Nuclear fuel, uranium dioxide, thermal conductivity, PWR. 1. Introduction With the evolution of the technology, theoretical and experimental researchers have been trying to streamline the empirical equations that predict the thermal conductivity of UO 2 , incorporating new models for the phonons and including the dependency of temperature in its functional form [1, 2]. Due to the low thermal conductivity of the fuel material (UO 2 ), a quite steep temperature gradient appears in the pellet. Considerably high temperatures are reached at the pellet center and an important safety criterion is to keep the temperature of the fuel below the melting point [3]. Corresponding author: Daniel Artur Pinheiro Palma, professor, research fields: the point kinetic equations, the Doppler broadening functions and fuel rod performance in nuclear power plants. E-mail: dapalma@cnen.gov.br. High-burnup is another interest of the nuclear power plants operators and researches about new models are desirable and can be found in the literature [4, 5]. In this paper the fthcon.f subroutine, contained in the MATPRO package of the FRAPCON-3.4 [6, 7] code was modified to include a new model proposed by Dias [8, 9]. The results obtained for temperature distribution in the center line of the fuel rod from the modified code were compared to those produced by the same code when using other models with an explicit dependency on temperature. 1.1 The FRAPCON 3.4 Code Description FRAPCON-3.4 is an analytical tool developed by PNNL (Pacific Northwest National Laboratory) that calculates LWR (light water reactor) fuel rod behavior D DAVID PUBLISHING