Development and Preliminary Verification of OpenMC-PARCS Two-step Criticality and Burnup Calculation Model for Fast Reactors

Fast reactors cannot directly use PWR calculation models for neutronics analysis due to their hard spectra and complex resonance phenomena. Monte Carlo (MC) method utilizes continuous-energy neutron cross-sections, which can accurately simulate resonance interference phenomena in fast reactors, yielding highly precise homogenized few-group cross-sections. This paper, based on MC method and the Triangle-based Polynomial Expansion Nodal (TPEN) method, investigates an OpenMC-PARCS two-step method of criticality and burnup calculation for fast reactors. Based on the OpenMC one-step method calculation results, a preliminary validation of the assumed constant microscopic cross-section burnup calculation scheme is conducted using the sodium-cooled fast reactor benchmark problem MET-1000. In the initial steady-state calculation, the deviation of the core effective multiplication factor (k_eff) using the OpenMC-PARCS two-step method is −104pcm (1pcm=10⁻⁵), and the deviation in the radial power distribution is no greater than 1%. During burnup calculations, the maximum deviation of the core k_eff from the reference solution is 591.2pcm, while most major nuclide number density deviation is no greater than 1%. The preliminary validation results indicate that the OpenMC-PARCS two-step method model can be used for large metallic fast reactor core design and fuel management.