Evaluation of the impact of the nuclear data library cinder.dat in MCNP burn-up calculations

Burn-up calculations with the MCNP code are based on the cinder.dat library, which includes the information of the cross-sections, the fission product yields and the decay data necessary to perform the depletion (among others). This library is based mainly on ENDF/B-VI.0 and has been enhanced with other databases, but no version based on other libraries (like JEFF or JENDL) is available in the bibliography or supplied with the MCNP code. This creates an inconsistency when other libraries are used for transport since the information of different libraries is mixed in the burn-up process.