Estimating the uncertainty in reactivity accident neutronic calculations

A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green's Function method