The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They are divided in the enrichment folder, where each salt has been parametrized over Li-6 enrichment fraction, and the mesh folder, where mesh results have been used for coming up with the cumulative curves over distance from the neutron source. Each folder contains a Python input as example of the model used
The goal of this study was to evaluate the impact of simulating different fuel shapes for the materi...
This release of OpenMC is the first release to use a new native HDF5 cross section format rather tha...
New Features Coarse mesh finite difference (CMFD) acceleration no longer requires PETSc Statepoi...
The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They...
OpenMC is part of the small modular reactor (SMR) simulation subproject "ExaSMR" in the broader Exas...
OpenMC, an open source Monte Carlo particle transport code, relies on its own nuclear data format th...
This work describes newly developed features of the OpenMC code for nuclear data processing. OpenMC,...
The input files of the spatially-uniform and non-uniform neutron sources used for the Monte Carlo si...
Combinatorial Geometry (CG) is one formulation for computational geometric models that is commonly u...
This release of OpenMC includes a few new major features including the capability to perform neutron...
The development of open-source applications in the nuclear field has recently attracted interest fro...
<p>Dataset used for the results presented in the CFD4NRS-9 conference paper "Simulation o...
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently ...
This release of OpenMC includes several new features, performance improvements, and bug fixes compar...
What's Changed Fix 2D mesh resolution by @RemDelaporteMathurin in https://github.com/fusion-energy/...
The goal of this study was to evaluate the impact of simulating different fuel shapes for the materi...
This release of OpenMC is the first release to use a new native HDF5 cross section format rather tha...
New Features Coarse mesh finite difference (CMFD) acceleration no longer requires PETSc Statepoi...
The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They...
OpenMC is part of the small modular reactor (SMR) simulation subproject "ExaSMR" in the broader Exas...
OpenMC, an open source Monte Carlo particle transport code, relies on its own nuclear data format th...
This work describes newly developed features of the OpenMC code for nuclear data processing. OpenMC,...
The input files of the spatially-uniform and non-uniform neutron sources used for the Monte Carlo si...
Combinatorial Geometry (CG) is one formulation for computational geometric models that is commonly u...
This release of OpenMC includes a few new major features including the capability to perform neutron...
The development of open-source applications in the nuclear field has recently attracted interest fro...
<p>Dataset used for the results presented in the CFD4NRS-9 conference paper "Simulation o...
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently ...
This release of OpenMC includes several new features, performance improvements, and bug fixes compar...
What's Changed Fix 2D mesh resolution by @RemDelaporteMathurin in https://github.com/fusion-energy/...
The goal of this study was to evaluate the impact of simulating different fuel shapes for the materi...
This release of OpenMC is the first release to use a new native HDF5 cross section format rather tha...
New Features Coarse mesh finite difference (CMFD) acceleration no longer requires PETSc Statepoi...