Abstract
Accurate evaluation of the neutron exposure of reactor structural materials is important either for preparing nuclear power plant life extension or decommissioning plan. Focusing on a typical BWR reactor model, this study presented a systematic comparison of three methods for determining the three-dimensional distribution of fast neutrons in the model. They were the approximate three-dimensional flux synthesis and two full three-dimensional transport calculations using TORT and MAVRIC, respectively. The flux synthesis was based on a one-dimensional ANISN and two two-dimensional DORT discrete ordinates solutions. TORT is a three-dimensional discrete ordinates transport code and MAVRIC is a hybrid deterministic/Monte Carlo transport code with continuous-energy capability. To facilitate comparisons between these versatile codes, the authors developed several utility scripts which speed the process of input preparation and output analyses. The comparisons among these methods was evaluated in pairs concerning their accuracies in geometry modeling and the predicted flux distributions as well as computational efficiencies.