A Generalized Multigroup Cross Section Generation Procedure for High-fidelity Reactor Simulation of Advanced Reactor Designs

Motivated by the increased importance of high-fidelity modeling capabilities that can improve the economics and safety of reactor designs, there are active movements to extend the applicability of high-fidelity neutronics codes to advanced reactors. The advanced reactor designs being developed by the industry are significantly diverse in fuel types, geometrical configurations, and neutron spectra. Many designs adopt particulate fuels and increasingly heterogeneous assembly and core geometries. Thermal, epi-thermal, and fast spectrum reactors are considered. This requires significant modifications and improvements in the current multigroup cross section generation methods and procedures, which are developed based on various assumptions and approximations tailored to a specific reactor type of interest. This necessitates it to develop a new generalized multigroup cross section generation procedure for more heterogeneous configurations and a wide range of spectrum. In this dissertation work, a generalized multigroup cross section generation method is developed for high-fidelity simulation of advanced reactors. In this method, important reactor characteristics over the entire energy range of fission reactors are properly modelled through the following approaches: (1) explicit modeling of resonance and anisotropic scattering of all isotopes by employing the detailed slowing down calculation, (2) proper consideration of spatial self-shielding for all regions, (3) development of an efficient lattice transport calculation method through a generalized condensation scheme, (4) preparation of problem-dependent cross section libraries with respect to background cross section and material temperatures, (5) development of a double heterogeneity treatment capability for particulate fuel modeling, and (6) generation of accurate gamma cross section library for coupled neutron and gamma heating calculations. The new multigroup cross section generation procedure in this work successfully yields substantially improved results for eigenvalue, reaction rates and heat generation rate for various benchmark problems regardless of reactor energy spectrum. This indicates that the developed procedure can improve the accuracy of high-fidelity neutronic calculations for advanced reactor analysis.PhDNuclear Engineering & Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttp://deepblue.lib.umich.edu/bitstream/2027.42/171338/1/bkjeon_1.pd