About this Research Topic
Monte Carlo (MC) transport methods have excelled in the field of particle transport simulation, owing to their superior modelling capabilities that enable the direct representation of complex simulation models using Computer-Aided Design (CAD). However, the precise definition of the entire universe is crucial for accurate simulations, which poses a significant challenge in converting Boundary Representation (BRep) representations into Boolean combinations of spaces suitable for Monte Carlo simulation in current nuclear analyses.
This Research Topic is dedicated to highlighting the recent advancements in CAD-to-MC modelling approaches, tool developments, and overcoming the difficulties often encountered during geometry preparation. The focus is on the application and verification of CAD-to-MC geometry modelling approaches for advanced nuclear systems (e.g., fission, fusion, and accelerator systems), whilst showcasing the achievements in the field.
The scope of this Research Topic encompasses three main CAD-to-MC methodologies:
• Constructive solid geometry
• Triangulation faceting
• Unstructured mesh
Topic themes of particular interest to this Research Topic include, but are not limited to, the following:
• Algorithm improvements (e.g. decomposition algorithms, void generation algorithms, faceting and meshing algorithms, and acceleration algorithms).
• Interface developments with popular simulation codes such as MCNP, OpenMC, Serpent, and Geant4, as well as user interface and auxiliary tool enhancements.
• Applications which demonstrate large-scale geometry modelling, with a particular emphasis on advances in fission nuclear systems, accelerator-driven neutron sources facilities, and fusion reactor systems.
All manuscript types are welcome in this Research Topic.
This Research Topic is dedicated to highlighting the recent advancements in CAD-to-MC modelling approaches, tool developments, and overcoming the difficulties often encountered during geometry preparation. The focus is on the application and verification of CAD-to-MC geometry modelling approaches for advanced nuclear systems (e.g., fission, fusion, and accelerator systems), whilst showcasing the achievements in the field.
The scope of this Research Topic encompasses three main CAD-to-MC methodologies:
• Constructive solid geometry
• Triangulation faceting
• Unstructured mesh
Topic themes of particular interest to this Research Topic include, but are not limited to, the following:
• Algorithm improvements (e.g. decomposition algorithms, void generation algorithms, faceting and meshing algorithms, and acceleration algorithms).
• Interface developments with popular simulation codes such as MCNP, OpenMC, Serpent, and Geant4, as well as user interface and auxiliary tool enhancements.
• Applications which demonstrate large-scale geometry modelling, with a particular emphasis on advances in fission nuclear systems, accelerator-driven neutron sources facilities, and fusion reactor systems.
All manuscript types are welcome in this Research Topic.
Keywords: Advanced Nuclear Systems, Fusion Reactor Systems, Fission Nuclear Systems, Accelerator-driven Neutron Sources Facilities, Monte Carlo, CAD, CSG, Modelling, Geometry, facet, unstructured mesh, MCNP
Important Note: All contributions to this Research Topic must be within the scope of the section and journal to which they are submitted, as defined in their mission statements. Frontiers reserves the right to guide an out-of-scope manuscript to a more suitable section or journal at any stage of peer review.
