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Welcome to Discourse
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0
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456
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August 5, 2020
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OpenMC monthly meeting dates 2024
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35
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1104
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May 27, 2025
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Online OpenMC Online Course June 30-July 3
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0
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7
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May 27, 2025
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How do i Install the latest version of openmc in my apple silicon chip Macbook Air M1
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3
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47
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May 27, 2025
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Invalid Tally Score: "ifp-time-numerator" - Unknown Score
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5
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39
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May 27, 2025
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Neutron flux in CEFR fuel assembly
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1
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26
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May 26, 2025
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Build openmc/openmc:develop for arm64
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0
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15
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May 24, 2025
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Job vacancies looking for OpenMC skills
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47
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3369
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May 23, 2025
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MGXS Fission equal to zero
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1
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23
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May 23, 2025
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Updating materials using depletion results
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0
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20
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May 23, 2025
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Evaluation determination of gamma scanning fuel burn up
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1
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23
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May 22, 2025
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95% of external source sites rejected
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2
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19
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May 22, 2025
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Papers using OpenMC
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193
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8095
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May 22, 2025
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Calculation of "Prompt neutron lifetime" and neutron generation time using Openmc
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1
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31
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May 22, 2025
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Pulse-height tally detector NaI(Tl)
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0
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14
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May 22, 2025
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Fission Locations of an SFR?
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0
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6
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May 21, 2025
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The geometry.xml of OpenMC can be calculated and run normally in the CPU version, but it cannot be run in the GPU version.
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1
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33
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May 21, 2025
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Modeling Xenon Transport in MSR
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0
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14
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May 21, 2025
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Photoneutron reactions
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5
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280
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May 20, 2025
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Material Density Variation
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3
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51
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May 20, 2025
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Transient with Control Drums
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1
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49
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May 20, 2025
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ERROR: No fission sites banked on MPI rank 0 for SMR PWR
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1
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20
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May 17, 2025
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How do i Replace the Nuclear Data Library in Openmc in Docker?
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3
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36
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May 17, 2025
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Failure in unit test
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11
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1381
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May 15, 2025
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Gamma detector simulation example
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1
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35
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May 14, 2025
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Plotting flux spectrum after depletion in openmc
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5
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188
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May 14, 2025
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WARNING: After particle 1712 crossed surface 69 it could not be located in any cell and it did not, ERROR: Maximum number of lost particles has been reached. leak.,,
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2
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16
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May 14, 2025
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Hexagonale forne
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1
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16
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May 14, 2025
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How to simulating series of materials without loading the cross-section data after every iteration?
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1
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28
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May 13, 2025
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Create homogenized cross sections for group of cells
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0
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22
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May 12, 2025
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