Integrated KENO Monte Carlo transport for 3-D depletion with scale

Accurate calculation of the depletion of nuclear materials requires careful determination of the neutron flux density and spectrum in the region(s) of interest. Increasing complexity in reactor designs, evolutionary concepts, and nonreactor applications such as safeguards, security, and nonproliferation, is beginning to require robust geometrical modeling capabilities to capture neutron transport for complex configurations. Monte Carlo transport methods offer the type of flexibility needed for such applications but will present other difficulties not encountered in deterministic transport methods. Two code sequences have been developed at Oak Ridge National Laboratory to perform Monte Carlo-based depletion. Both the KENO V.a and KENO-VI Monte Carlo transport codes in the SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system have been adapted to work within a depletion code sequence; comparisons with other methods and measured data show excellent agreement.