Abstract
ARTI (Advanced Radiation Technology Institute, KAERI) has been operating a 30 MeV cyclotron and is going to produce $^{89}Zr$ which is useful for diagnosis of cancer as an alternative of FDG (FluoroDeoxyGlucose). It is produced by proton irradiation on $^{89}Y$ target. Shielding analysis of a transport container has not been carried out, though there has been few cases of producing $^{89}Zr$ in Korea. In this study, we carry out shielding analysis for the $^{89}Zr$ transport container with MCNPX code. $^{89}Zr$ of which physical half life is 78.4 hours emits 0.909 MeV energy through a gamma transition at 99.9% fraction and ${\beta}^+$ emission at 23% fraction. The material of transport container is 97% lead with 3% antimony alloy, and the shape is cylindrical with diameter of 85 mm, height of 124 mm and weight about 8 kg. Radiation sources in transport container and were modeled based on the above data. Detectors which is a cylindrical shape with a radius and a height of 1cm respectively were set up from the outside of the transport container. Total of 8 tallies defined as F4 tally were also applied to calculate of flux for volume. Calculation results using MCNPX show that transport container can contain $^{89}Zr$ up to approximately 40 mCi to meet the radiation regulation of nuclear Act. This calculation is based on the surface radiation dose rate of $2mSv{\cdot}h^{-1}$ given in the radiation regulations. However, since the conditions used in analyses are conservative, actual maximum activity would be higher.
