Evaluation of the FRANCO finite element fuel rod analysis code

Abstract

Knowledge of the temperature distribution in a nuclear fuel rod is required to predict the thermal and mechanical response of the strongly temperature-dependent fuel elements. In this research, the FRANCO ( F inite Element Fuel R od AN alysis CO de) computer code was developed for use on an IBM-PC to predict thermal and mechanical fuel performance characteristics for reactor operating conditions. A cross-sectional area of a fuel rod is discretized using constant strain triangular elements. To simplify the time-dependent fuel performance analysis, this code uses quasi-steady state conditions or slow power changes for time-independent problems; thus, time-dependent cases can be represented by quasi-static “snap shots.” The development and testing of the FRANCO program for deformation and heat transfer problems is described. The fuel performance models and solution procedures are provided in detail. The accuracy of FRANCO is examined through selected benchmark problems with other nuclear industry finite difference and finite element fuel analysis codes and actual Halden experimental measurements. These benchmarks show that FRANCO performs well except for the fuel-clad contact cases when compared to results from other fuel performance codes, such as FREY and ESCORE. The FRANCO program developed in this research effort is easy to run, fast, and yields comparable results.