Solving the neutron transport equation on unstructured mesh and its application on high-fidelity multi-physics analysis of micro reactors

摘要: High-fidelity unstructured mesh based neutron transport method is necessary to accurately model the behavior of micro reactors considering high neutron leakage, the thermal-hydraulics and geometry deformation feedback effects. Based on the MOOSE framework, the multi-group self-adjoint angular flux neutron transport equation is solved. The unstructured mesh modeling capability provides flexibility in representing irregular shapes and boundaries in micro reactors, while the combination of finite element and discrete ordinate methods ensures efficient and accurate solution of the neutron transport equation. The neutron transport solver was then applied for the neutronics analysis of Xi’an Pulsed Reactor, steady-state multi-physics analytical benchmark problem and coupled neutronics/heat transfer/thermal expansion analysis of the Godiva prompt critical transient. The ability of treating the nonlinearity including source-term coupling between different physics, material nonlinearity and geometric nonlinearity are proved in our analysis. Future work will be conducted on more realistic micro reactor designs like heat pipe cooled reactors and gas cooled reactors.