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About the User Support category
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0
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532
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August 13, 2020
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Candle gfr with uranium metallic
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0
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9
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May 26, 2023
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Did not sample any reaction for nuclide U235
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10
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58
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May 26, 2023
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Openmc 13.3 depletion results is not as expected
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9
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198
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May 26, 2023
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Polyethylene (HDPE) cross section for neutron moderation
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8
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53
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May 25, 2023
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Openmc Voxel plot not showing in peraview
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4
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135
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May 25, 2023
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Finding absorption and fission cross-section from openmc for a simple 1d geometry
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1
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14
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May 24, 2023
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Heating from Neutron Interaction
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6
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42
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May 23, 2023
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Geometry Error using openmc.Quadric
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4
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29
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May 23, 2023
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OpenMC quits using universe.plot
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0
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25
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May 22, 2023
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Installing from Source
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2
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58
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May 20, 2023
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Determine TRISO Locations
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4
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62
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May 20, 2023
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Python API issues
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3
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40
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May 19, 2023
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Debugging a simple photon transport model generated with ChatGPT
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4
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268
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May 18, 2023
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TRISO particle outside of lattice
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3
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165
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May 18, 2023
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H-3 in spent fuel pellets
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3
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49
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May 18, 2023
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Save and load weight window as xml
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0
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17
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May 18, 2023
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Tests that use their own CMake
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0
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22
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May 17, 2023
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Modification to StatePoint Tally Values
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0
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26
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May 17, 2023
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A CellFromFilter for Materials
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0
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19
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May 17, 2023
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Creating simple wall geometry which is set in x axis but infinite in y and z axis
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0
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14
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May 16, 2023
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Problem with chain file using OpenMC through Docker
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5
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51
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May 16, 2023
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Openmc.lib open pseudoterminals
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0
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17
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May 16, 2023
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Mat should be of type openmc.Material or str?
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1
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29
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May 16, 2023
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Problem with photon-neutron transport
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8
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68
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May 16, 2023
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GFR with Uranium Metalic (CANDLE)
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0
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23
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May 14, 2023
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Uranium carbide benchmark
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0
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18
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May 13, 2023
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Difference between OpenMC and Serpent keff results
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3
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133
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May 12, 2023
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Tally Uncertainty; OpenMC vs. MCNP
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12
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446
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May 11, 2023
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Question for surface_source_write/read
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2
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33
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May 10, 2023
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