Workshop on Criticality Calculations with MCNP

NRSC’s staff member participated in a workshop which took place in Los Alamos National Laboratory; the workshop was devoted to Criticality Calculations with the MCNP. The workshop mainly was focused on best MCNP application practices as how to perform criticality calculations for reactor physics and criticality safety applications by applying the MCNP. During the workshop the following topics were mainly addressed:

  • criticality calculations by using Monte Carlo methods,
  • geometry, including lattices and repeated structures,
  • tallies and mesh tallies,
  • cross-section data,
  • statistical analysis,
  • assessment of the convergence of Keff and the source distribution (Shannon Entropy),
  • 3D depletion capabilities of MCNP6, and
  • sensitivity analysis of Keff.

NRSC uses MCNP to carry out independent analysis of criticality safety of dry spent fuel storage, the ANPP spent fuel pools and spent fuel transport cask for Armenian Nuclear Regulatory Authority’s decision-making. Furthermore, recently the NRSC started to apply the MCNP for 3D depletion analysis of WWER-440 spent fuel assemblies for further burnup credit, severe accident and radiological consequence analysis.

3D Depletion Model of WWER-440 Fuel Assembly developed by MCNP6