Validation Study of Pin Heat Transfer for UO_2 Fuel Based on the IFA-432 Experiments

摘要

The Integrated Fuel Assessment IFA-432 experiments from the International Fuel Performance Experiments database were designed to study the effects of gap size, fuel density, and fuel densification on fuel centerline temperature in light water reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for uranium dioxide (UO_2)fuel systems was performed, with a focus on the densification stage (2.2 GWd/tonne UO_2). In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole. The analysis demonstrated excellent agreement for rods 1, 2, 3, and 5 (varying gap thicknesses and density with traditional fuel), demonstrating the accuracy of the codes and their underlying material models for traditional fuel. For rod 6, which contained unstable fuel that densified an order of magnitude more than traditional, stable fuel, the magnitude of densification was overpredicted, and the temperatures were outside the experimental uncertainty. The radial power shape within the fuel was shown to have a significant impact on the predicted centerline temperatures, whereas the effect of modeling the fuel at the thermocouple location as either annular or solid was relatively negligible. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for UO_2 fuel with respect to a well-validated nuclear fuel performance code.