Comparative Study of Neutronics Analysis Techniques for Radioactive Waste Assessment

Abstract

Computational models created for neutronics assessment through solid geometry conversion are often specific to the analysis being performed. The use of unstructured mesh geometry has the potential to reduce the build time of MCNP models, reduce inaccuracies introduced through flux averaging over different components and material mixing, and make use of computer-aided design models that can also be suitable for other types of analysis. In this paper three neutronics methods were investigated for suitability in performing a radioactive waste assessment of a fusion demonstration reactor. The methods included the conventional cell-based approach, a superimposed structured mesh, and the use of a recently developed capability with unstructured mesh geometry. It was concluded that an unstructured mesh approach has the potential to be an important tool for assessing radioactive waste to inform reactor and component design.