Preceeding of PHYSOR 2012, Knoxville, Tennessee, USA, April 15-20, 2012, on CD-ROM
The purpose of this paper is to quantify uncertainties of fuel pin cell or fuel assembly (FA)
homogenized few group diffusion theory constants generated from the B1 theory-augmented
Monte Carlo (MC) method. A mathematical formulation of the first kind is presented to quantify
uncertainties of the few group constants in terms of the two major sources of the MC method;
statistical and nuclear cross section and nuclide number density input data uncertainties. The
formulation is incorporated into the Seoul National University MC code McCARD. It is then used
to compute the uncertainties of the burnup-dependent homogenized two group constants of a lowenriched UO2 fuel pin cell and a PWR FA on the condition that nuclear cross section input data of
U-235 and U-238 from JENDL 3.3 library and nuclide number densities from the solution to fuel
depletion equations have uncertainties. The contribution of the MC input data uncertainties to the
uncertainties of the two group constants of the two fuel systems is separated from that of the
statistical uncertainties. The utilities of uncertainty quantifications are then discussed from the
standpoints of safety analysis of existing power reactors, development of new fuel or reactor
system design, and improvement of covariance files of the evaluated nuclear data libraries.
