Abstract

We have developed a hybrid Unstructured Mesh (UM)/Constructive Solid Geometry (CSG) model of the next-generation spallation neutron target-moderator-reflector-shield (TMRS) assembly Mark-IV at the Los Alamos Neutron Science Center (LANSCE) to perform MCNP neutronics calculations in support of the engineering design process. The hybrid model consists of an essential component of the new TMRS—the upper target—built as UM and the rest of the model that resembles the original TMRS Mark-III built as CSG. The UM model for the neutronics analysis as well as the model for the subsequent engineering analyses such as structural stress and fluid dynamics are automatically converted from the identical computer aided design (CAD) model. Such an automatic conversion is highly time-efficient and preserves a great level of detail of the original model. In addition, only a relatively small UM portion of our model (0.003 m3) is embedded into the surrounding CSG universe (314 m3), enabling a more efficient MCNP calculation due to the reduced requirements on computer memory and shorter computation time in comparison with a fully extended UM model. Our MCNP calculations focus on the neutronics performance and energy deposition in the upper target. They employ high-energy nuclear physics models, tabular data, and scattering kernels, use an 800 MeV proton beam as a source, tally secondary fast and thermal neutrons with point detectors, and apply mesh tallies and elemental edit outputs to score energy deposition from all particles, i.e., neutrons, photons, protons, and charged pions. Because the use of UM marks a major departure from our original method based entirely on the traditional CSG modeling with MCNPX 2.7.0, we validate the new hybrid UM/CSG model against the legacy CSG model. We found a good agreement between the two models, which demonstrates that MCNP6.2 with UM/CSG hybrids can solve complex neutronics problems at large accelerator facilities, including spallation neutron sources.

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