Thermal and stress analysis of a composite nuclear fuel rod using nonlinear finite element techniques

This work presents the development of a nonlinear, axisymmetric, finite element code and details its use in the steady state thermal stress analysis of a composite nuclear fuel rod. The active length of the fuel rod is composed of UO2-PuO2 fuel pellets, incorporating a central void subject to radiation heat transfer, interspersed with molybdenum washers to enhance conductive heat transfer from the center of the rod to the cladding surface. The finite element code, written to solve general problems in axisymmetric thermo-elasticity, is capable of handling nonlinear material properties as well as such diverse boundary conditions as enclosure radiation, gap conductance phenomena, and frictional coupling between surfaces. The results of the code applied to the analysis of the composite fuel rod show a maximum fuel temperature, at a linear power level of 18 Kw/ft (with 10000F sodium coolant) of 39700F, 6000F less than comparable conventional fuel designs and some 1400F less than predicted by earlier work which did not take into account center void radiation heat transfer. Thermal stresses, while not accurately predictable using an elastic axisymmetric code due to the high temperatures and stress levels involved and the resultant fracturing of the fuel pellet, were shown to be lower than the equivalent stresses in conventional fuel rod designs operating under the same conditions