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About the Validation and Verification category
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0
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586
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August 13, 2020
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Volume Calculation and Keff discrepancies SERPENT /OPENMC
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0
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41
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June 15, 2025
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Pulse-height tally detector NaI(Tl)
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0
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22
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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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10
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May 21, 2025
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Difference in multiplication factor
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1
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245
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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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17
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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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111
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March 16, 2025
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Multiple transfer rate
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0
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20
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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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96
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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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51
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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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47
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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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56
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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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64
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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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96
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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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223
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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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134
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August 22, 2024
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Neutron Inelastic Scattering
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2
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64
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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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70
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August 7, 2024
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Continuing to investigate convergence
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0
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47
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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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78
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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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215
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June 11, 2024
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OpenMC-MCNP heating validation
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0
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126
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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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149
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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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196
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March 1, 2024
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Plotting Geometry Problem
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2
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322
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February 28, 2024
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Error in cross-section
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6
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231
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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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264
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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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247
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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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178
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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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120
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February 15, 2024
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