DAGMC slice plotter rotating geometry?
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6
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156
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September 29, 2023
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Neutron tracking
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5
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178
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September 13, 2023
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Decay heat is quite low
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4
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179
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January 9, 2024
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Problem locating OpenMC via mamba
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2
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250
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November 6, 2023
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[Need help] Keff results significantly different from SCALE6.1
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2
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241
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September 5, 2023
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How to calculate the neutron flux of each lattiece universe (corresponding cell)
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6
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161
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November 30, 2023
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paramak.ExtrudeCircleShape not working?
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6
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155
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August 24, 2023
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Parallel computation for openmc.deplete
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2
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238
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December 16, 2023
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Asking for a solution to a criticality problem
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5
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176
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October 17, 2023
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Is there a specific tutorial for OpenMC to run Distributed-Memory Parallelism on Ubuntu
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4
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183
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November 3, 2023
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Need to understand photon interactions
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4
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166
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September 7, 2023
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How to get the jendl5-hdf5 crosssection library for multiple temperature?
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3
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196
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August 15, 2023
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"Volume not specified for material ID=2" even after assigning volume to mix material
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5
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159
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November 2, 2023
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Fetching Depletion Results
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0
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63
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March 29, 2024
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Kernel crash when run integrator.integrator()
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4
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173
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October 4, 2023
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OSError: statepoint file has a version of 18.1 which is not consistent with the version expected by OpenMC, 17
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2
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260
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July 6, 2023
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No tally output when using UnstructuredMesh
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5
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165
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July 4, 2023
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No cross_sections.xml file was specified in materials.xml or in the OPENMC_CROSS_SECTIONS
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2
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240
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August 11, 2023
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MGXS library generation with dagmc geometry
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8
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141
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April 30, 2024
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ERROR: More than 95% of external source sites sampled were rejected. Please check your external source’s spatial definition
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2
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239
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August 23, 2023
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Depletion Results vs TBR in fusion source
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5
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169
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October 24, 2023
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Pymoab error with openmc-dagmc-wrapper
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4
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178
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August 16, 2023
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Inelastic cross section (n,level)
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2
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126
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March 19, 2024
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Standard debugging not giving more info than RuntimeError: OpenMC aborted unexpectedly
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3
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207
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December 7, 2023
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Setting depletion power equal 0 to perform decay calculations stopped working
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3
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177
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December 20, 2023
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Geometry plotting not working as expected
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3
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202
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July 21, 2023
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OpenMC within Docker
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3
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177
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July 10, 2023
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Precision of openmc.deplete.Results.get_atoms()
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1
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281
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July 15, 2023
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Neutron flux and reaction rates discrepancies with MCNP
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2
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209
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January 19, 2024
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Volume Calculation Taking a Long Time
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8
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123
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February 21, 2024
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Difference between heating and heating-local tallies
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1
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256
|
August 7, 2023
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Identical Result
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3
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178
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July 26, 2023
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Problem with U235 concentration with burnup of the PWR fuel cell
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2
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210
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February 27, 2024
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V 13.3 Setting of volume in cells manually doesn't work with .get_nuclides_densities
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4
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164
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June 23, 2023
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OpenMC 0.14.0 Shared-Memory Parallelism not working
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4
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148
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December 1, 2023
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Error with 'diff_burnable_mats' in depletion of fuel assembly
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8
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124
|
March 31, 2024
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Intermediate and Advanced OpenMC Workshops
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1
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231
|
September 25, 2023
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Error executing command: "pip install ."
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4
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157
|
October 23, 2023
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Installing old versions
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8
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121
|
January 26, 2024
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Cannot define region from hexagonal lattice boundary to a cylinder
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2
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204
|
June 26, 2023
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Spherical Source, no interaction
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5
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142
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October 3, 2023
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Error converting tally results to vtk file
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5
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137
|
September 1, 2023
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Krusty core design and neutronics
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2
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199
|
September 18, 2023
|
How do you rotate this model by 180/90 degrees
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4
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164
|
August 29, 2023
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The energy spectra of neutron point sources calculated by mcnp and openmc are inconsistent
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4
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163
|
July 31, 2023
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TRISO particles in a hexagonal fuel lattice
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5
|
128
|
June 23, 2023
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Keff indiscrepency when modelling TRISO fuel particles
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1
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117
|
March 1, 2024
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OpenMC eigenvalue simulation coupled with burnup calculation considering online reprocessing for a Molten Salt Reactor
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1
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243
|
September 4, 2023
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CAD geometry compilation error
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5
|
152
|
August 29, 2023
|
Tallying alongside depletion
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2
|
202
|
August 19, 2023
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