Impact of thermal conductivity models on the coupling of heat transport, oxygen diffusion, and deformation in (U, Pu) O _(2-X) nuclear fuel elements

摘要

We study the coupled thermal transport, oxygen diffusion, and thermal expansion in a generic nuclear fuel rod consisting of a (U_(1-y)Pu _y)O_2-x) fuel pellet separated by a helium gap from zircaloy cladding. Steady-state and time-dependent finite-element simulations with a variety of initial- and boundary-value conditions are used to study the effect of the Pu content, y, and deviation from stoichiometry, x, on the temperature and deformation profiles in this fuel element. We find that the equilibrium radial temperature and deformation profiles are most sensitive to x at small values of y. For larger values of y, the effects of oxygen and Pu content are equally important. Following a change in the heat-generation rate, the centerline temperature, the radial deformation of the fuel pellet, and the centerline deviation from stoichiometry track each other closely in (U,Pu)O _(2-x), as the characteristic time scales of the heat transport and oxygen diffusion are similar. This result is different from the situation observed in the case of UO_(2+x) fuels.