5.5.1 Introduction

The aim of this part is to present some reactor simulation examples; it cannot replace the MCNP reference manual.

When MCNP is running, two special files are needed; one is, of course, the problem description (geometry, materials, neutron source, etc.). The other one, named xsdir, is a special file that contains some nuclear data and the path of the ENDF library files for each material. This file can be placed in a default directory, but MCNP first looks in the ‘current’ directory if such a file exists.

5.5.2 Units

The specific units used in MCNP are summarized in table 5.1.