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About the Validation and Verification category
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0
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580
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August 13, 2020
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Difference in multiplication factor
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1
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230
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May 9, 2025
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Geometry of fuel log TMSR-500 ThorCon International design
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0
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15
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May 2, 2025
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Converting OpenMC Flux Unit to Match MCNP
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3
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89
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March 16, 2025
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Multiple transfer rate
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0
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19
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March 3, 2025
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Trying to model SD-TMSR using OpenMC
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3
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87
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December 7, 2024
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ERROR get_atoms from depletion result
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1
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42
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December 4, 2024
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Astonishing axial zone divsion effects
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2
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41
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December 4, 2024
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Zircaloy-4 activation with JEFF33 depletion chain (Nb97_m1)
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2
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50
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November 8, 2024
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Neutron spectrum of MSRE cad design
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0
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61
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October 16, 2024
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The difference flux between openMC and MCNP(MCX)
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2
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84
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October 11, 2024
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Difference in neutron flux between OPENMC and MCNP6
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1
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194
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September 24, 2024
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Transmutation results different from FISPACT-II
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4
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123
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August 22, 2024
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Neutron Inelastic Scattering
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2
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57
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August 12, 2024
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Error while running a Pincell model
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4
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63
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August 7, 2024
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Continuing to investigate convergence
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0
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45
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July 31, 2024
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K-inf natural Uranium infinite geometry keff seems too low
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1
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74
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July 10, 2024
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Depletion problem with multiplying media in external sources
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5
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193
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June 11, 2024
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OpenMC-MCNP heating validation
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0
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121
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May 8, 2024
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Keff converges, but the value depends on the initial seed
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1
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144
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April 21, 2024
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Keff indiscrepency when modelling TRISO fuel particles
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1
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190
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March 1, 2024
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Plotting Geometry Problem
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2
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305
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February 28, 2024
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Error in cross-section
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6
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211
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February 27, 2024
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Problem with U235 concentration with burnup of the PWR fuel cell
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2
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261
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February 27, 2024
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Serpent and OpenMC TRISO Discrepency
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0
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236
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February 20, 2024
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Serpent vs Open MC Discrepancy for PinCell (TRISO Fuel)
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0
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167
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February 20, 2024
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MCNP5 Error with Thermal Cross Sections
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0
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119
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February 15, 2024
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Neutron flux and reaction rates discrepancies with MCNP
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2
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376
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January 19, 2024
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Sodium Fast Reactor Benchmarks
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7
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1037
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December 8, 2023
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OpenMC for modeling and simulating SMRs
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1
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279
|
October 29, 2023
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