Potential Impacts of Modeling Full Reactor Cores Using Combined Fuel Performance and Thermal Hydraulics Codes

Nuclear reactor systems codes have the ability to model the system response in an accident scenario based on known initial conditions (ICs) at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermomechanical fuel rod response models needed for best-estimate prediction of fuel rod failure. Alternatively, the reverse can be said about fuel performance codes; they can lack the ability to capture and model the thermal-hydraulic (T-H) influence of adjacent fuel rods and the rod's location in the reactor core. This work analyzes the limitations in using fuel performance codes to represent in-reactor conditions as determined by full-core T-H codes. The codes used in this analysis are the U.S. Nuclear Regulatory Commission's steady-state fuel performance code FRAPCON-3.5 and T-H code TRACE-V5P3. In order to assess the impact of the limitations found in the codes, several modifications were made to all of the codes to improve code-to-code consistency. The modifications to the fuel performance code include adding the ability to model gamma-ray heating and providing realistic core coolant conditions. The T-H code modifications include adding the ability to model the fuel with axially varying burnup-dependent fuel and cladding dimensional changes and corrosion characteristics. The fuel in a Westinghouse four-loop pressurized water reactor was modeled to assess the impacts these modifications have on fuel performance and ICs for transient analysis. The results of this study show that current modeling assumptions (and limitations) can yield both conservative and nonconservative results on several important licensing criteria.