International Conference on Mathematics, Computional Methods & Reactor Physics 2009, CD-ROM.
The method for analyzing uncertainty propagation in Monte Carlo (MC) depletion calculations
implemented in the McCARD code is examined with a depletion calculation for a TRU fuel
assembly. The effects of uncertainties in the nuclear cross section data and in the nuclide number
density on the k∞ prediction and also on the depletion behavior of the actinides are quantified
with the use of JENDL 3.3 covariance data. It is shown that the cross section uncertainties are the
major contributors to the uncertainties of the neutronic parameters of the FA. By splitting the
uncertainty in k∞ into several individual components originating from each individual cross
section uncertainty, it is identified that which types of nuclear data need be improved in order to
obtain a more reliable nuclear design of the TRU FA.
