S.N. Cerezo, S.S. Chirayath, G. Long, C. Keith, “Stochastic And Data Uncertainty Quantificaiton Of MCNP Predicted Nuclide Concentrations In Fuel Burnup Simulations”, INMM/ESARDA Joint Annual Meeting, 22-26 May 2023, Vienna, Austria.
Monte Carlo N-Particle transport code (MCNP) is used widely 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, which are useful in nuclear safeguards monitoring. During fuel burnup simulations, the uncertainties in the predicted nuclide concentration due to the uncertainty in the nuclear data and the stochastic nature of the MCNP methodology are not propagated. The nuclide concentration is calculated through CINDER 90 isotope generation and depletion module in MCNP which uses the neutron reaction rates and flux values. Stochastic uncertainties in the neutron reaction rates and flux values are calculated by MCNP, which introduces stochastic uncertainty in the nuclide concentration, but this uncertainty is not propagated through each time step. This uncertainty is in addition to the systematic uncertainty due to the nuclear data. The reaction rates can be broken down into flux, number density, and microscopic cross section terms. The number density and flux are provided by MCNP, and the flux term will contain the stochastic uncertainty; however, the microscopic cross sections are not provided by MCNP and will contain a systematic uncertainty. Propagating the effects of both sources of uncertainty using a Backward Euler numerical scheme allows for the reporting of the total relative error in the predicted nuclide concentrations. The MCNP depletion calculation uses a one group flux and therefore a one group microscopic cross section is necessary to find the neutron reaction rates. The microscopic cross sections are dependent on the energy spectrum of the flux in the fuel burnup simulation. The process of acquiring these microscopic cross sections and weighting them by the flux is automated in our study for estimating both stochastic and systematic uncertainties. The final product of the study will be a software that automatically calculates and reports stochastic, systematic, and total nuclide concentration uncertainties from a single MCNP output file.