Parameterized representation of macroscopic cross section in burn-up cycles

Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized macroscopic cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density. In this study is evaluated the problem-dependent about fission, scattering and capture cross sections for a typical PWR fuel element reactor, considering burn-up cycles. The analysis was carried out with the SCALE 6.1 code package. Tests realized as the temperature coefficient of reactivity, fast fission factor, and the comparison with direct calculations with the SCALE code system and with Lagrange polynomials show excellent agreements. The differences between the cross section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03%.