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About the User Support category
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0
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37
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August 13, 2020
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How to obtained the multi-group homogenized cross-sections using OpenMC
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2
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14
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December 3, 2020
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Geometry can change during depletion calculation?
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3
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46
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December 3, 2020
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How to specify the boron concentration (ppm) in calculation
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3
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29
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December 2, 2020
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Thermal Scattering Data in 300,900 and1800 kelvin
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4
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42
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November 23, 2020
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Source convergence problem
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4
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34
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November 23, 2020
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Error while running the same model for different particles per batch
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2
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54
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November 23, 2020
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Question about M301 and M901 in non-fissile nuclides
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1
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18
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November 23, 2020
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Question about power designation while depleting
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5
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45
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November 22, 2020
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Create materials.xml from depletion results?
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8
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94
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November 18, 2020
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Exporting HDF5 file
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3
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43
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November 17, 2020
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Mismatch between OpenMC documentation and latest development deplete.Operator arguments sintaxis
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1
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30
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November 17, 2020
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openmc.model.RectangularParallelepiped usage
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4
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40
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November 17, 2020
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Depletion question
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5
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60
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November 17, 2020
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Continue-run
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3
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30
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November 17, 2020
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How to convert the time step of OpenMC 2D depletion calculation from day/second to burn-up depth (MWth/TU)
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2
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27
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November 13, 2020
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Question on ADS geometry
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1
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45
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November 11, 2020
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CMake Error(could not find HDF5)
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3
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36
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November 6, 2020
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How to indicate cross section data
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25
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123
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November 5, 2020
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Is there any simple example for serrated reactor core
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8
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64
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November 4, 2020
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'openmc' returned non-zero exit status 255
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15
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50
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November 4, 2020
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Tally current across assembly surfaces of a hexagonal core
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3
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18
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November 4, 2020
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Negative volume value
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2
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26
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November 2, 2020
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OpenMC setup Google Colab
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4
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46
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November 1, 2020
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Energy filters of neutrons in Flux calculation
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3
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23
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October 28, 2020
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Nuclear data dependent zero result for heating-local tally
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1
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18
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October 27, 2020
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How to calculate the pin power distribution
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2
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45
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October 27, 2020
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Adjoint Mode possible in OpenMC?
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1
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22
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October 26, 2020
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Thermal Scattering Data
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13
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57
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October 23, 2020
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Depletion Chain File with ENDF-B-7.0
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8
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80
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October 22, 2020
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