A NEW FACILITY FOR MCNP APPLICATION IN WHOLE BODY COUNTING AND INTERNAL DOSIMETRY

Development of precision measurement techniques (such as whole body spectrometry), as well as techniques for processing the results of such measurements, opens an opportunity to increase essentially the reliability of internal irradiation monitoring of nuclear industry personnel. The major difficulties of such monitoring are related to measurements of low-energy -radiation which is intensively absorbed and scattered in body tissues (e.g., for such biologically important radionuclides, as 239 Pu and 241 Am). The attenuation of radiation depends on individual anatomy of a patient. Such individual anatomy cannot be taken into account correctly when using plastic phantoms (mannequins) for calibration of whole body spectrometers. On the contrary, calculations can provide adequate representation of the individual if radiation transport is simulated in voxel geometry retrieved from medical tomography images. For such calculations, only Monte Carlo method can be used in practice. The paper reviews a recently developed graphical user interface for automatic creation of MCNP input data files. The voxel geometry of radiation transport is restored on the base of the patient’s tomograms. The new software enables also graphical analysis of calculation results. The GUI was successfully used for whole body counter calibration as well for studies of influence of individual anatomical features on counting efficiency of whole body spectrometers. The accurate description of the patients anatomy provided by the utility allows correction of interpretation of the measured date by at least of a two-fold factor. The new technique seems rather important for radiation monitoring of nuclear industry personnel.