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Discrepancy Between Legendre and Mesh Flux Tally Magnitudes
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0
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22
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September 4, 2024
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Can the OpenMOC program build fuel plate elements?
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0
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13
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September 4, 2024
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Modelling Am241/Be Source
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1
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57
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September 1, 2024
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Questions about MultiGroup databases
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0
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10
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September 1, 2024
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Easiest Method to Copying Cells
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3
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32
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August 31, 2024
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About data processing
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0
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17
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August 30, 2024
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Starting depletion calculation from a previous simulation
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1
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27
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August 30, 2024
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Hexahedral lattice
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0
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20
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August 29, 2024
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Question regarding cross section data
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0
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25
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August 29, 2024
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"Output for a MOAB mesh was requested" error
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0
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10
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August 28, 2024
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Setting an array to "UniverseBase" data type issue
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3
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19
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August 27, 2024
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Asking for help to fix
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3
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26
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August 26, 2024
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AttributeError: 'tuple' object has no attribute '__name__'
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1
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7
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August 23, 2024
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Simulation gets "stuck" between batches
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7
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39
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August 24, 2024
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Reaction rate and Flux normalization problems
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2
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54
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August 22, 2024
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How to set tally for the rotated lattice universe?
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0
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30
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August 21, 2024
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Mass change in depletion runs
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2
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35
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August 20, 2024
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Error in Destination Material
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2
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20
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August 20, 2024
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Non-isotropic source and custom source problem
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0
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31
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August 18, 2024
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About constant power normalization during eigenvalue calculation and depletion and photo-fission contribution to the eigenvalue
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3
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60
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August 17, 2024
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Photoneutrons in Beryllium Reflected Reactors
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2
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27
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August 16, 2024
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OpenMC - OpenFOAM coupling
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0
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37
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August 16, 2024
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Generate_single_ww_and_apply.ipynb is not working with fission neutrons.Neutrons do not promot any further
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4
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36
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August 16, 2024
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Simulation left hanging on the addition of a mesh tally
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0
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18
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August 16, 2024
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Issue with Depletion Calculations
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4
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51
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August 16, 2024
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Universe plotting: RuntimeError: Particle -1 left lattice
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2
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22
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August 16, 2024
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Computing (effective) delayed neutron fraction
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6
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1321
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August 15, 2024
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Two or more universes are not used as fill universes
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1
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16
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August 14, 2024
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Geometry problem
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12
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75
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August 14, 2024
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Pincell model error while attempting to calulate keff
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3
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14
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August 14, 2024
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