Preliminary Comparison of Transport Codes Applied to a Second-Generation PT-SCWR Lattice

Abstract

The Pressure-Tube Supercritical Water Cooled Reactor (PT-SCWR) is being developed in Canada as part of the Generation IV International Forum (GIF) efforts to develop future nuclear-power systems. The PT-SCWR consists of a vertical, non-pressurized, cylindrical vessel filled with heavy-water moderator. The vessel contains a rectangular array of fuel channels oriented axially, each consisting of a pressure tube and an inner ceramic thermal insulator and containing 78-element Th-Pu fuel bundles. A two-dimensional lattice benchmark has been previously developed to assess the applicability of various lattice physics codes to the PT-SCWR design. This work summarizes the benchmark results for two lattice codes: Serpent, a stochastic (Monte Carlo) transport code, and DRAGON, a deterministic transport code. Specifically, reactivity, select reactivity coefficients, and multi-group macroscopic cross-sections as a function of burnup are compared. Preliminary results show a 4mk difference in the reactivity values and a maximum 6% difference in two-group macroscopic cross sections.