Abstract

The MCNP code was used for simulation of the GC1518 HPGe gamma spectrometer at the Center for Nuclear Techniques Ho Chi Minh City. The input data of geometrical dimensions and material compositions of gamma spectrometer were set up based on practical measurement and data supplied by the manufacturer. These input data were checked by comparison of calculated and experimental gamma spectra of 137Cs and 60Co reference sources. The agreement between calculation and experiment results was obtained for two source potions inside the lead shield. This shows that the input data were accurate and acceptable for the use of the MCNP code in other caculations.