1099 Transactions of the American Nuclear Society, Vol. 116, San Francisco, California, June 11–15, 2017 Reactor Analysis Methods—I Thermal Analysis of a Fuel Channel Abdullah Weiss, Xue Yang Department of Mechanical and Industrial Engineering, Texas A&M University-Kingsville, 700 University Blvd. / MSC 191, Kingsville, TX 78363 USA, abdullah.weiss@students.tamuk.edu INTRODUCTION To better study and understand the efficiency of fuel channels in reactors, a thermal analysis method was developed to conduct a study on any fuel channel to gain a better understanding of the temperature, heat rate, heat flux, and power profiles of such fuel channels. This study is conducted on a typical PWR (Pressurized Water Reactor) fuel assembly. The justification for specifically choosing a PWR fuel channel is because PWRs are the most widely used type of reactors around the world. This study was performed using NEWT code in SCALE 6, ANSYS, and SolidWorks. MATLAB and Excel were also utilized for some of the calculations. The power and heat rate of the fuel rod were calculated, alongside a 2D mapping of temperature, heat flux, directional heat flux, and thermal error on a 3D model of the fuel channel by analyzing the fuel rod and the surrounding water moderator independently. METHODS The study was done by evaluating the relative power of a fuel assembly in the core using 2D Transport: T- NEWT code in SCALE 6 [1]. The NEWT input file consisted of a typical PWR assembly [2], represented in a rectangular array lattice (Fig. 1(a)) consisting of multiple fuel channels (Fig. 1(b)). Then by calculations done in Excel and MATLAB, the power, heat rate, and temperature of each fuel rod in the assembly is tabulated, and the average axial temperature distribution in a fuel rod is estimated. Then, the temperature distribution was plotted on a SolidWorks 3D model of the fuel pin using ANSYS, where the heat flux, directional heat flux, and thermal error in the fuel and the surrounding water was solved. Fig. 1(a). Fuel Lattice Fig. 1(b). Fuel Channel Power and Heat Rate Calculation The NEWT output file provides the relative fission power of different regions in the lattice, whereby the fuel rods can be identified as the only regions with an actual relative fission power value greater than zero. Using these values, the power of each fuel rod can be calculated. First, find the power in each fuel assembly using the thermal power of a typical PWR, which is 3411 MWTh, and the number of fuel assemblies in a typical PWR (193 assemblies): The relative fission power of each fuel rod in the NEWT output file is then multiplied with the power per assembly value found above, and the values obtained represent the total power of each fuel rod. Taking the average of all these power values allows for the calculation of the axial heat rates (Equation 1) by integrating the given equation [2]: (1) Where is a constant, reference heat rate value in the equation, integrating yields the equation for power (Equation 2), , which can be rearranged to find the value for . (2) 1. Average fuel rod power ( ) 2. quation constant ( ) 3. Length of fuel rod ( ) 4. Axial location on the fuel rod ( ), which is estimated to be halfway through the fuel rod (at 1.8 m). Plugging in the values yields , which is used to find the axial heat rate (Equation 1) and power distribution (Equation 2). Temperature Distribution Calculation To obtain the axial temperature distribution, the thermal resistances (Fig. 2) in the fuel channel were considered in order to calculate appropriate values at any axial location on the fuel rod.