Calculation methods for multigroup cross sections used in burnup studies

Calculation of burnup of material composing a nuclear reactor is important in a feasibility study of lifetime of fuel in possible reactor systems. This work discusses (i) a basis for calculating a flux spectrum in order to obtain multigroup cross sections and (ii) a method for solving the burnup equations. Under (i) particular attention is paid to the calculation of multigroup resonance cross sections and (ii) a simple and effective analytic technique is derived. In addition, this work presents, in a consistent manner, developments in the field of reactor physics which effect the calculation of multigroup data. The physical model adopted to simplify the otherwise unmanageable equations is appropriate to a large recirculating fuel reactor operating under equilibrium conditions.