Development of thermal hydraulic correlations for the University of Texas at Austin TRIGA reactor using computational fluid dynamics and in-core measurements
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Safety is a paramount concern in the operation of training and test reactors. A major component of a reactor is the maintenance of safe thermal hydraulic operating conditions. If the temperature of the water coolant exceeds the boiling point, the heat transfer out of the fuel rods into the coolant will greatly decrease and will need to rely upon other safety feedbacks and systems to avoid an accident condition. TRIGA thermal hydraulic systems are currently modeled using a finite differencing code, TRACE/SNAP, developed by the Nuclear Regulatory Commission. While the code is currently certified, it has shortcomings that this work improves upon, notably the simplification of the more complex flow geometries by using circular pipes and a heat transfer correlation that is valid across all flow regimes observed during operation of the TRIGA. A computational fluid dynamics code, FLUENT, along with real-time thermocouple probe measurements of the channel were used to solve both of these major issues. A high resolution model of four adjacent flow channels was created to provide a numerical experimental data set for enhancing the correlations used in the TRACE model. The hot flow channel is connected to three surrounding channels where crossflow occurs causing a more complex flow pattern than the isolated single channel system used in TRACE/SNAP. To calibrate the FLUENT model, a thermocouple probe was designed and placed in the TRIGA core in the center of the flow channel. The reactor was operated over the full range of licensed power levels to obtain a fully encompassing data set of coolant temperatures. The FLUENT model was then adjusted so that the temperatures at the location of the probe in the model matched those from the experimental measurements. Based on the results from the FLUENT testing, data was extracted to develop a new heat transfer coefficient correlation and loss factor coefficient correlation due to the non-circular geometry and fuel rod end fittings for use in the TRACE/SNAP code. These adjustments were then implemented into TRACE/SNAP to improve the code for future users performing safety analysis on TRIGA reactors.