A single-step framework to generate spatially self-shielded multi-group cross sections from Monte Carlo transport simulations. (March 2019)

• Record Type: Journal Article
• Authors: Boyd, William; Forget, Benoit; Smith, Kord
• Is Part Of: Annals of nuclear energy. Volume 125(2019)
• Page Start: 261
• Page End: 271
• Publication Date: 2019-03
• Subjects: Neutron transport -- Multi-group cross-sections -- Spatial self-shielding -- Spatial homogenization; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2018.11.017
• Languages: English
• ISSNs: 0306-4549

Abstract

Abstract: A key challenge for full-core transport methods is reactor agnostic multi-group cross section (MGXS) generation. Monte Carlo (MC) presents the most accurate method for MGXS generation since it does not require any approximations to the neutron flux. This paper introduces a framework that replaces the traditional multi-level approach to MGXS generation with a single MC calculation to generate the fine-spatial mesh MGXS that are needed by high-fidelity transport codes. In addition, three pin-wise spatial homogenization schemes are introduced to model inter-pin spatial self-shielding spectral effects with varying levels of spatial fidelity. Most notably, the Local Neighbor Symmetry (LNS) scheme uses a nearest neighbor-like analysis of a reactor geometry to determine which fuel pins should be assigned the same MGXS. The single-step MGXS framework with the three spatial homogenization schemes are evaluated for two PWR benchmarks. The results demonstrate the potential for single-step MC simulations of the complete heterogeneous geometry as a means to generate reactor agnostic MGXS for deterministic transport codes.