About this Research Topic
The field of CAD-to-MC geometry modelling approaches for advanced nuclear systems is rapidly evolving, driven by the need for precise and efficient particle transport simulations. Monte Carlo (MC) transport methods have become a cornerstone in this domain due to their ability to model complex systems using Computer-Aided Design (CAD) accurately. However, a significant challenge persists in converting Boundary Representation (BRep) models into Boolean combinations of spaces suitable for MC simulations. This conversion is crucial for reliable nuclear analyses.
Recent studies have made strides in developing tools and methodologies to address these challenges, yet gaps remain in the seamless integration and verification of these models, particularly in advanced nuclear systems such as fission, fusion, and accelerator systems. Ongoing research centres around optimizing these conversions to enhance simulation accuracy and efficiency, highlighting the need for further investigation and innovation in this field.
This Research Topic aims to explore and highlight advances in CAD-to-MC geometry modelling approaches specifically tailored for advanced nuclear systems. Key areas of interest include improving the accuracy of BRep to Boolean space conversions, enhancing the integration of these models with existing simulation codes, and verifying the application of these methodologies in real-world scenarios.
To gather further insights into CAD-to-MC geometry modelling for advanced nuclear systems, we welcome articles addressing, but not limited to, the following themes:
• Algorithm improvements, including decomposition, void generation, faceting, meshing, and acceleration algorithms
• Interface developments with popular simulation codes such as MCNP, OpenMC, Serpent, and Geant4, alongside user interface and auxiliary tool enhancements
• Applications demonstrating large-scale geometry modelling, with a focus on advances in fission nuclear systems, accelerator-driven neutron source facilities, and fusion reactor systems.
Recent studies have made strides in developing tools and methodologies to address these challenges, yet gaps remain in the seamless integration and verification of these models, particularly in advanced nuclear systems such as fission, fusion, and accelerator systems. Ongoing research centres around optimizing these conversions to enhance simulation accuracy and efficiency, highlighting the need for further investigation and innovation in this field.
This Research Topic aims to explore and highlight advances in CAD-to-MC geometry modelling approaches specifically tailored for advanced nuclear systems. Key areas of interest include improving the accuracy of BRep to Boolean space conversions, enhancing the integration of these models with existing simulation codes, and verifying the application of these methodologies in real-world scenarios.
To gather further insights into CAD-to-MC geometry modelling for advanced nuclear systems, we welcome articles addressing, but not limited to, the following themes:
• Algorithm improvements, including decomposition, void generation, faceting, meshing, and acceleration algorithms
• Interface developments with popular simulation codes such as MCNP, OpenMC, Serpent, and Geant4, alongside user interface and auxiliary tool enhancements
• Applications demonstrating large-scale geometry modelling, with a focus on advances in fission nuclear systems, accelerator-driven neutron source facilities, and fusion reactor systems.
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.
