A new methodology to estimate stochastic uncertainty of MCNP-predicted isotope concentrations in nuclear fuel burnup simulations

The Monte Carlo N-particle radiation transport code (MCNP) is used widely in nuclear fuel burnup simulation (FBS). In the FBS, MCNP predicts various neutron reaction rates and the associated stochastic relative error (SRE). The SRE is due to the nature of the statistical methods used in MCNP. These SREs for neutron reaction rates and fluxes are reported in the MCNP output for each FBS time-step. However, MCNP does not report SREs for its predictions of isotope concentrations. A new methodology to compute the SRE for the MCNP-predicted isotope concentrations is developed, which uses MCNP-computed neutron reaction rates and the associated SREs. The effectiveness of the methodology is verified for a pressurized water reactor FBS and is found to be satisfactory. The SRE computation for isotope concentrations in irradiated fuel has applications in nuclear science and engineering.