MCNP를 이용한 Zr-89 운반용기의 차폐적정성 평가
Evaluation of Shielding Capability of 89Zr Transport Container using MCNP Code
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다운로드
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참고문헌
키워드
초록
ARTI (Advanced Radiation Technology Institute, KAERI) has been operating a30 MeV cyclotron and is going to produce 89Zr which is useful for diagnosis of cancer as analternative of FDG (FluoroDeoxyGlucose). It is produced by proton irradiation on 89Y target. Shielding analysis of a transport container has not been carried out, though there has beenfew cases of producing 89Zr in Korea. In this study, we carry out shielding analysis for the 89Zrtransport container with MCNPX code. 89Zr of which physical half life is 78.4 hours emits 0.909MeV energy through a gamma transition at 99.9% fraction and β+ emission at 23% fraction. Thematerial of transport container is 97% lead with 3% antimony alloy, and the shape is cylindricalwith diameter of 85 mm, height of 124 mm and weight about 8 kg. Radiation sources in transportcontainer and were modeled based on the above data. Detectors which is a cylindrical shape witha 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. Calculationresults using MCNPX show that transport container can contain 89Zr up to approximately 40 mCito meet the radiation regulation of nuclear Act. This calculation is based on the surface radiationdose rate of 2 mSv·h-1 given in the radiation regulations. However, since the conditions used inanalyses are conservative, actual maximum activity would be higher.
Abstract
ARTI (Advanced Radiation Technology Institute, KAERI) has been operating a30 MeV cyclotron and is going to produce 89Zr which is useful for diagnosis of cancer as analternative of FDG (FluoroDeoxyGlucose). It is produced by proton irradiation on 89Y target. Shielding analysis of a transport container has not been carried out, though there has beenfew cases of producing 89Zr in Korea. In this study, we carry out shielding analysis for the 89Zrtransport container with MCNPX code. 89Zr of which physical half life is 78.4 hours emits 0.909MeV energy through a gamma transition at 99.9% fraction and β+ emission at 23% fraction. Thematerial of transport container is 97% lead with 3% antimony alloy, and the shape is cylindricalwith diameter of 85 mm, height of 124 mm and weight about 8 kg. Radiation sources in transportcontainer and were modeled based on the above data. Detectors which is a cylindrical shape witha 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. Calculationresults using MCNPX show that transport container can contain 89Zr up to approximately 40 mCito meet the radiation regulation of nuclear Act. This calculation is based on the surface radiationdose rate of 2 mSv·h-1 given in the radiation regulations. However, since the conditions used inanalyses are conservative, actual maximum activity would be higher.
참고문헌 (7)
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