Adjustment of group cross sections by means of integral data (ENDF/-VII.1)

The purpose of neutronic calculations is to determine many principal integral parameters such as effective multiplication factor (keff), reaction rate, spectrum indices, etc. These parameters, are based on several cross sections as well as their uncertainties. However, the uncertainty propagations effect will give, in turn, inaccurate values of these integral parameters. Therefore, the margin of reactor safety can be decreases. In order to minimize these risks, the adjustment of basic nuclear data (Cross section) is required.