Assessment of TRITON and PARCS for Full-Core MOX Fuel Calculations

摘要

The neutron physics of MOX fuels may represent a challenge to existing computational tools. However, the more rigorous treatment of neutron transport available within NEWT, coupled with the accuracy of ORIGEN-S depletion capabilities and SCALE self-shielding calculations within TRITON-driven lattice physics analyses, provides a rigorous first-principles approach for calculation of cross sections for such fuel designs. Although not initially developed to be a general production tool for MOX analyses, the TRITON sequence provides a rigorous physics treatment of MOX fuel transport and depletion, providing improved insight into the physics of such systems. It is anticipated that this knowledge will allow improvement in other analysis methods and/or data to help meet the needs of future MOX-fueled cores.