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Easy way to compute solid angle of neutron source
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0
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13
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May 12, 2025
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MCNP to OpenMC tool
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15
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2427
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May 10, 2025
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Hexagonal assembly
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2
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45
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May 9, 2025
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Difference in multiplication factor
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1
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237
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May 9, 2025
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Normalization Factor
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0
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22
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May 8, 2025
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Negative photon heating
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1
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32
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May 7, 2025
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RectangularPrism not of type region
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3
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24
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May 7, 2025
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Getting scatter probability from tally score
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0
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13
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May 7, 2025
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Volume for depletion
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1
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36
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May 6, 2025
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The different kinf between MCNP and OpenMC
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1
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252
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April 30, 2024
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TMSR-500 fuel logs design of slab and slot position in one fuel log,
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0
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15
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May 2, 2025
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Material cannot reflect in lower universe and Particle 17551 could not be located after crossing a boundary of lattice 3
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2
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21
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May 2, 2025
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Geometry of fuel log TMSR-500 ThorCon International design
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0
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16
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May 2, 2025
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Inconsistent flux tallies between random ray and monte carlo eigenvalue simulation
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3
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58
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May 2, 2025
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Criticality Search
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3
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57
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April 30, 2025
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Simulation freezes when using weight windows (version 0.14)
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6
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228
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April 30, 2025
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Source Rates for depletion Integrators
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0
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29
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April 29, 2025
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Different distribution of Neutron density in openmc_plasma_source
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7
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199
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April 25, 2025
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Bremsstrahlung Dose Tally with DAGMC Geometry — Extremely High Dose Values
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0
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22
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April 24, 2025
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Determing real simulation time
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6
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82
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April 23, 2025
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Depletion/Burnup not implementing into example problem
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5
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64
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April 23, 2025
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X- ray Exposure room simulation
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1
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33
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April 23, 2025
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Depletion calculations do not deplete
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11
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831
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April 23, 2025
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Keff remains unchanged in depletion calculation
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9
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386
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April 23, 2025
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Problem with reflective surface
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4
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61
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April 22, 2025
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Remove_redundant_surfaces() for identical planes with different a, b, c, d (proportional)
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2
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34
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April 21, 2025
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Reaction frequency using multigroup cross-section and inverse-velocity
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0
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17
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April 19, 2025
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Get_decay_heat(), no depletion chain file?
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2
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37
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April 17, 2025
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Natural abundances of Tantalum
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1
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31
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April 16, 2025
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Is possible to separate tally result according to source distribution values?
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2
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26
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April 15, 2025
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