Development of multigroup cross section generation code MC2-3 for fast reactor analysis

Under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of U.S. DOE, an integrated, advanced neutronics code system that allows the high fidelity description of nuclear reactors and simplifies the multi-step design process is being developed. As part of this effort, an advanced multigroup cross section generation code named MC2-3 is being developed for fast reactor applications by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating these improved methods with the two-dimensional method of characteristics solver of the high-fidelity transport code UNIC. In order to enhance accuracy and efficiency, various improvements have been made on the ETOE-2/MC2-2 code system that has been successfully used for decades in generating multigroup cross sections for fast reactor analysis. First of all, in order to eliminate the limitations of the current generalized resonance integral method for resolved resonance self-shielding, a new approach based on the numerical integration of pointwise resonance cross sections with the narrow resonance approximation has been introduced. The continuous slowing-down method used for the spectrum calculation in the resolved resonance energy range has been replaced by the consistent P1 method with extended transport approximation up to P9 in order to minimize the approximations involved in light elements (in particular, hydrogen) treatment and to simplify the program structure.The consistent P1 equations are solved in ultra-fine (∼2,100) group level; the hyper-fine (>50,000) group spectrum calculation in the resolved resonance range can optionally be invoked to enhance the resolved resonance self-shielding accuracy. A new capability of handling anisotropic inelastic scattering matrices has also been added. These improved methods and one-dimensional transport calculation capability have been integrated into a new cell code MC2-3 with a modern programming structure. The new cell code has been incorporated into the neutron transport code UNIC to generate the multigroup cross sections consistently with the material and temperature distributions used in transport calculations. The cell calculations provide the multigroup cross sections for two-dimensional lattice or whole-core transport calculations, depending on the desirable level of approximation appropriate for the computational resources and analysis goals of the user. Initial verification tests of the MC2-3 code and its ENDF/B-VII.0 libraries have been performed using various fast critical experiments including eight LANL critical assemblies, ZPR-6/6A ZPR-6/7, ZPPR-15, and ZPPR-21 critical experiments. The resulting effective multiplication factors have shown very good agreements with MCNP5 and VIM Monte Carlo solutions: within 150 pcm for the LANL critical experiments; within 40 pcm for ZPR-6 assembly 6A and 7; within 250 pcm for six configurations of ZPPR-21 and three configurations of ZPPR-15A