Nuclear Science and Engineering / Volume 178 / Number 2 / October 2014 / Pages 172-185
Technical Paper / dx.doi.org/10.13182/NSE13-19
The IFA-597 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments database were designed to study the thermal behavior of mixed oxide (MOX) fuel and the effects of an annulus on fission gas release in light water reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for MOX fuel systems was performed, with a focus on the first 20 time steps (∼6 GWd/tonne initial heavy metal) for explicit comparison between the codes. 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, dish, and chamfer. The analysis demonstrated relative agreement for both solid (rod 1) and annular (rod 2) fuel in the experiment, demonstrating the accuracy of the codes and their underlying material models for MOX fuel, while also revealing a small energy loss artifact in how gap conductance is currently handled in Exnihilo for chamfered fuel pellets. The within-pellet power shape was shown to impact the predicted centerline temperatures significantly. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for MOX fuel with respect to a well-validated nuclear fuel performance code. This analysis was done for a heavy-water boiling water reactor, but the conclusions are not limited by the reactor type beyond the spectrum and particular materials.