Citation:
S.N. Cerezo, “Systematic Uncertainty Quantification Of MCNP Predicted Nuclide Concentrations In Fuel Burnup Simulations”, M.S. Thesis, Nuclear Engineering, Texas A&M University, College Station, TX (2024).
Abstract:
Monte Carlo N-Particle transport code (MCNP) is often used to simulate nuclear fuel burnup and depletion because it is efficient in solving the radiation transport equation for complex geometries. MCNP simulates fuel burnup and estimates the concentrations of actinides and fission products generated in the fuel, which are useful in nuclear forensics as well as safeguards monitoring. During fuel burnup simulations, the uncertainties in the predicted nuclide concentrations due to the uncertainty in the nuclear data used by MCNP are not propagated and predicted. The nuclide concentration is calculated through CINDER 90 isotope generation and depletion module in MCNP. The CINDER90 module uses the neutron reaction rates and flux values computed by MCNP for each burnup time step. The reaction rates can be broken down into three terms: neutron flux, number density of the target isotope that is transmuting, and microscopic neutron interaction cross section. The number density and neutron flux are provided by MCNP; however, the microscopic cross sections are not directly provided by MCNP in the output and will contain systematic uncertainty in varying degrees depending on the microscopic cross section of the target isotope of interest. Systematic uncertainty is not propagated through each MCNP burnup time step. Propagating the effects of systematic uncertainty using a Backward Euler numerical scheme allows for the reporting of the systematic relative error in the predicted nuclide concentrations, which the study undertaken in this thesis. This Backward Euler methodology was executed through python scripting and a program was developed to output the systematic relative error for user desired isotopes of interest utilizing on the results of MCNP fuel burn up simulation. It was concluded that the Backward Euler methodology and the Bateman equations successfully replicated the MCNP estimated concentrations given the appropriate one group cross sections. Additionally, it was determined that for select isotopes of interest the systematic uncertainty for the associated concentration can be estimated.