Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.Cataloged from PDF version of thesis.Includes bibliographical references (pages 55-56).OpenMC is an open source Monte Carlo code designed at MIT with a focus on parallel scalability for large nuclear reactor simulations. The target problem for OpenMC is a full core high-fidelity multi-physics coupled simulation. This encompasses not only nuclear physics, but also material science and thermohydraulics. One of the challenges associated with this problem is efficient data management, as the memory required for tallies alone can easily enter the Terabyte range. This thesis presents an efficient system for data storage which allows for physic...
New Features Meshline plotting capability Support for plotting cells/materials on middle universe ...
The present work describes the latest advances and progress in the development of the OpenMC Monte C...
OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. T...
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently ...
A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uran...
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, ...
An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is devel...
Performance results are presented for a multi-threaded version of the OpenMC Monte Carlo neutronics ...
OpenMC is part of the small modular reactor (SMR) simulation subproject "ExaSMR" in the broader Exas...
Modern scientific datasets present numerous data management and analysis challenges. State-of-the- a...
Monte Carlo neutron transport codes are a growing subject of research in nuclear reactor analysis. F...
One of the challenges of high granularity calorimeters, such as that to be built to cover the endcap...
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 20...
Many relationships between parameters and physical properties in materials science and engineering a...
The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They...
New Features Meshline plotting capability Support for plotting cells/materials on middle universe ...
The present work describes the latest advances and progress in the development of the OpenMC Monte C...
OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. T...
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently ...
A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uran...
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, ...
An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is devel...
Performance results are presented for a multi-threaded version of the OpenMC Monte Carlo neutronics ...
OpenMC is part of the small modular reactor (SMR) simulation subproject "ExaSMR" in the broader Exas...
Modern scientific datasets present numerous data management and analysis challenges. State-of-the- a...
Monte Carlo neutron transport codes are a growing subject of research in nuclear reactor analysis. F...
One of the challenges of high granularity calorimeters, such as that to be built to cover the endcap...
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 20...
Many relationships between parameters and physical properties in materials science and engineering a...
The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They...
New Features Meshline plotting capability Support for plotting cells/materials on middle universe ...
The present work describes the latest advances and progress in the development of the OpenMC Monte C...
OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. T...