Effects of Source and Geometry Modeling on the Shielding Calculations for a Spent Nuclear Fuel Dry Storage Cask

Chen, Y. F.; Chiou, Yi-Fu; Chang, Soon-Bok; Jiang, Shiang‐Huei; Sheu, Rong-Jiun

doi:10.13182/nt13-a16432

Cited by 3 publications

(1 citation statement)

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“…Monte Carlo codes have been used to calculate the radiation level outside the FCTC10 container with the maximum loading [4]. Only steel shells, tungsten alloy and lead fillings were considered in the calculation.…”

Section: Radiation Level 1) Calculationsmentioning

confidence: 99%

Measurements and Assessments on Shielding Performance of FCTC10 <sup>60</sup>Co Transport Container

Zhuang

Zhang

et al. 2016

J Radiat Prot Res

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Background: FCTC10 container is designed to transport 60 Co radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a 60 Co source of 1.8 × 10 5 Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

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Section: Radiation Level 1) Calculationsmentioning

confidence: 99%