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Depletion/Transmutation Decay Chain Clarification
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2
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86
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August 27, 2025
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How to get the JENDL5.HDF5 crosssections?
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2
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114
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February 21, 2025
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Help to create CFER
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2
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89
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March 26, 2025
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Python API not found using conda installation
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4
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52
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October 18, 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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87
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May 22, 2025
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Issue Integrating Decay Gamma Data from ENDF/B-VIII.0 into OpenMC Full Chain
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3
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65
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August 7, 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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85
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May 17, 2025
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Lower Bound of Weight Window
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2
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92
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December 5, 2025
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RuntimeError:No fission sites banked on MPI rank 0
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2
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71
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April 8, 2025
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Plotting cylindrical mesh flux
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2
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79
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February 25, 2025
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95% of external source sites rejected
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2
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69
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May 22, 2025
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MPI Multi-Node Calculations - SLURM
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8
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50
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January 20, 2026
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Plot cross section
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1
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76
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June 13, 2025
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Composite materials - okay to add S(alpha, beta) cross sections?
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1
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75
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March 18, 2025
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Extending a universe in one direction
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4
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63
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August 13, 2025
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DAGMC Module Not Found in Python After Docker Build with OpenMC and DAGMC Support
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4
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59
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June 28, 2025
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Libopenmc.so Error
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3
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55
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July 3, 2025
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Normalization when power is 0
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3
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61
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March 25, 2025
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Simulation of photonuclear reaction in coupled neutron-gamma simulation
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1
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87
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March 5, 2025
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Heating and Flux score affecting each other
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0
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31
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December 23, 2025
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Get_decay_heat(), no depletion chain file?
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2
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62
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April 17, 2025
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Adjustable Orientation for Right Circular Cylinder
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1
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80
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March 27, 2025
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K-effective fluctuation with beryllium nuclide addition in higher than room temperature
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3
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57
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March 11, 2025
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Normalisation Procedure (Mesh Volume)
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1
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78
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August 28, 2025
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Using the openmc.VolumeCalculation
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2
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62
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April 4, 2025
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About openmc.data.njoy.make_ace_thermal
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2
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67
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July 15, 2025
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Energy Deposition With Photon Transport
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1
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74
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June 12, 2025
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Photon Transport for Radioisotope Source
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1
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70
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March 29, 2025
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Problem on the openmc.run() ,cross sections file
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1
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78
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February 19, 2025
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Pin-wise flux tally in SFR hexagonal assembly (one hex duct cell per fuel pin)
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1
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41
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December 24, 2025
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Openmc Room Simulation for X-Ray
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2
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50
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June 15, 2025
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Creating Cross Section Files at 959K and 900K
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2
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61
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March 13, 2025
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Possible Issue in Decay-Only Simulation with Fixed-Source Transmutation
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3
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57
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October 28, 2025
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Burnup and Reaction rate
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1
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67
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November 15, 2025
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What is the energy distribution of a default fixed source
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1
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66
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July 15, 2025
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Depletion analysis on 3D cad geometry
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1
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71
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April 11, 2025
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Depletion Chains
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1
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64
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March 26, 2025
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Volume Calculation and Keff discrepancies SERPENT /OPENMC
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0
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92
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June 15, 2025
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Control on the standard deviation
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1
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63
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July 7, 2025
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My CMFD isn't working properly
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1
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63
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June 6, 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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52
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May 14, 2025
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SolidRayTracePlot -> empty plot (aka white square)
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4
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44
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March 18, 2025
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Accessing data while looping through multiple runs
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3
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49
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March 26, 2025
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Neutron flux in CEFR fuel assembly
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1
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63
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May 26, 2025
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Need help. Failed to reproduce the tutorial "csg_to_cad" of Cardinal
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0
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17
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January 23, 2026
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Control blade rotation
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2
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58
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March 1, 2025
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Depletion Arguments for a BWR
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2
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59
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February 14, 2025
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openmc.deplete.Resultats - material difference
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1
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60
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October 31, 2025
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Final Year Project Help: Troubleshooting a TMSR500 Fuel Log Geometry in OpenMC
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1
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59
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October 24, 2025
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Fixed source depletion
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1
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62
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August 20, 2025
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