SARAX: A new code for fast reactor analysis part I: Methods

A renaissance in fast reactor research and development and the rapid development of computing technologies have encouraged improvements in fast reactor modeling and calculation. A new code named SARAX has been developed. This paper reviewed current fast reactor neutronics analysis codes and provided an overview of SARAX developments. We introduced the following three methods: a hybrid method for cross-section generation, a two-level coarse mesh method for transport calculation acceleration, and a combined sensitivity and uncertainty analysis method. We discussed preliminary tests to show the benefits of these methods from two perspectives: (1) replacing the flux solver from neutron diffusion to neutron transport and (2) using a hybrid method rather than using the all Monte Carlo method to generate the cross-sections. We discussed an error cancelation phenomenon based on the numerical benchmarks.