INITIAL VERIFICATION OF THE HIGH-FIDELITY NEUTRON TRANSPORT CODE PROTEUS FOR HETEROGENEOUS GEOMETRY PROBLEMS

摘要

The unstructured mesh-based neutron transport code PROTEUS was developed at Argonne National Laboratory under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to provide a heterogeneous geometry neutron transport capability for coupled multi-physics nuclear reactor applications. As a code verification effort, two 2D heterogeneous geometry reactor problems, the Advanced Test Reactor (ATR) and the C5 benchmark problem based on the C5G7 OECD benchmark, are analyzed. Eigenvalue and flux solutions of PROTEUS are compared with multi-group Monte Carlo (MCNP) solutions using the same 23 multi-group cross sections in order to eliminate the source of error arising from the heterogeneous cross sections themselves. This procedure allows the solver itself to be verified. Agreement is within 299 pcm Δk for the ATR eigenvalue and within 4.6% for fuel plate fluxes. For the C5 core, eigenvalues from both codes agree within 125 pcm Δk and show excellent agreement for pin cell fluxes. This work verifies the accuracy of PROTEUS given a sufficiently refined finite element mesh and angular cubature. The next phase of this work will focus on code verification with heterogeneous cross section generation.