Leakage-corrected discontinuity factors for a second-generation Thu-Pu pressure-tube SCWR

The neutron flux throughout a reactor core should be calculated ideally, by solving the neutron transport equation for a highly detailed geometric model of the core. Since this is computationally impractical, approximate node-homogenized models have historically been used whereby neutronic properties are averaged over cartesian parallelepipedic regions called nodes. This process is referred to homogonization. This simplest homogenization procedure is known as standard homogenization. Standard homogenization calculates node-homogenized cross sections as flux-weighted averages over the volume of each node. It uses an approximate spatial flux distribution obtained from single-node detailed geometry calculations that approximate the node-boundary conditions to be reflective. While standard homogenization has been successfully used for CANDU reactors, there exist more advanced homogenization methods such as Generalized Equivalence Theory (GET). GET improves accuracy by allowing the neutron flux in the node-homogenized model to be discontinuous at node boundaries through the use of discontinuity factors. Node-averaged cross sections and discontinuity factors can be obtained from single-node calculations using reflective boundary conditions. To further improve accuracy, non-reflective boundary conditions that approximate the real node-boundary conditions can be used; as process known as leakage correction. This work explores the use of GET with leakage-corrected cross sections and discontinuity factors for the next-generation PT-SCWR flux calculations. Results show that using GET in conjunction with leakage corrections yields substantial improvements in accuracy over standard homogenization and should be given serious consideration as a method for performing neutronic calculations for PT-SCWR cores. (author)