Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN are directly coupled with a fuel rod temperature module. Starting with a zero-dimensional burnup code such as ORIGEN, the importance of one- and more-dimensional neutron flux calculations in the field of isotope generation and depletion calculation is shown. OREST results are compared with measured isotope concentrations of depleted uranium dioxide samples and of mixed oxide (MOX) rods irradiated in different assembly positions. In addition, published results from two lattice cell burnup program systems are shown. Currently, ORIGEN (1973 version) is applied by many users in a stand-alone mode. The achievable accuracies are discussed. Only a few measurements of irradiated MOX fuels have been available. Considering the actual projects for reprocessing and recycling of nuclear fuels, further and fully documented isotope analyses are needed.
