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학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
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Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.
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참고문헌 (2)
참고문헌 신청
Oak Ridge National Laboratory / 2010 / OrigenArp Primer: How to Perform Isotropic Depletion and Decay Calculations with SCALE/ORIGEN / ORNL/TM-2010/43 :21~36
Los Alamos National Laboratory / 2003 / MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, Volume I. LA-UR-03-1987. Chapter 2. 2003;1-139. Volume II. LA-CP-03-0245. Chapter 3 :1~36
Los Alamos National Laboratory / 2003 / MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, Volume I. LA-UR-03-1987. Chapter 2. 2003;1-139. Volume II. LA-CP-03-0245. Chapter 3 :1~36
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