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How to calculate the Keff value when run in fixed source mode?
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16
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1821
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December 16, 2025
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Importance Game
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1
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34
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December 15, 2025
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Issues with loading previously generated weight windows
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3
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51
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December 15, 2025
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OpenMC on CUDA?
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9
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928
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December 13, 2025
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Pulling beta decay data from ENDF files`
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0
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27
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December 11, 2025
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Burnup, Onion skin effect and Gd problem
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0
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37
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December 8, 2025
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Calculating Neutron Thermal Flux Spectrum in TRIGA Reactor
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0
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31
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December 8, 2025
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Lower Bound of Weight Window
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2
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92
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December 5, 2025
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How to define a source in the shape of an octagon?
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0
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28
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December 4, 2025
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Total power due to leakage
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0
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19
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December 3, 2025
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PWR Lattice Modeling square vs hexagonal bundles
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0
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31
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December 2, 2025
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Control Drum Not Showing in the Plot File
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2
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46
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December 2, 2025
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Broaded detector response
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2
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41
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December 2, 2025
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The shared fission bank is full
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1
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232
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November 30, 2025
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Installing OpenMC on Apple Silicone using conda
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0
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41
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November 26, 2025
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I tried to reproduce the literature results, but failed
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0
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56
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November 24, 2025
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How to tally gamma production in neutron simulations?
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0
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26
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November 21, 2025
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Mono-Energetic Isotropic Surface Source Functionality
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2
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51
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November 19, 2025
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Burnup and Reaction rate
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1
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67
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November 15, 2025
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Validation of the Two-step Approach
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2
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98
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November 15, 2025
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ENDF -> HDF5 conversion support
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6
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162
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November 13, 2025
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TRISO Model XML Generation and HPC Simulation with different Parameters
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7
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103
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November 13, 2025
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Problems when using a cylindrical mesh in a fixed source simulation
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1
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46
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November 12, 2025
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Sphinx documentation error during pull request
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0
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20
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November 11, 2025
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Python API installation
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5
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117
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November 10, 2025
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(probably stupid) question about current tally backscattering
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3
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87
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November 7, 2025
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Did not sample any nuclide during collision
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12
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760
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November 7, 2025
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RuntimeError: std::bad_alloc after failed depletion on all other openmc runs!
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0
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31
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November 6, 2025
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Setting the maximum neutron transport energy
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0
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29
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November 6, 2025
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Depletion Chain File decay_energy Query
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0
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37
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November 4, 2025
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