Calculations for TRIGA ACPR with MCNPX

The purpose of the paper is to present an MCNP model for the TRIGA pulsed reactor (ACPR) operated by INR-Pitesti. MCNPX is a Fortran90 Monte Carlo radiation transport computer code that can be used in a broad range of applications for neutrons (and for a series of other particles, too) on a continuous (or discrete) energy domain. The reactor model created has been tested in different calculations, by comparing the bank control rods efficiency against experimental measurements, the prompt neutrons generation time and the effective fractions of delayed neutrons with the Safety Analysis Report parameters. Also, neutron flux in pulsed mode inside the experimental cavity have been calculated using the model and compared with results of flux measurements. Generally, the MCNPX model is used in static criticality calculations, to determine the reactor kinetics parameters and in the design of experiments. (authors)