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Home / Publications / Journals / Nuclear Technology / Volume 175 / Number 1 / Pages 343-350

Surface Dose Rate Calculations of a Spent-Fuel Storage Cask by Using MAVRIC and Its Comparison with SAS4 and MCNP

Y. F. Chen, R. J. Sheu, S. H. Jiang, J. N. Wang, U. T. Lin

Nuclear Technology / Volume 175 / Number 1 / July 2011 / Pages 343-350

Technical Paper / Special Issue on the 16th Biennial Topical Meeting of the Radiation Protection and Shielding Division / Radiation Transport and Protection

Based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology, MAVRIC is a new computational sequence in the SCALE6 code package that is designed to perform efficient Monte Carlo simulation for a complicated and difficult shielding problem. This study aimed to evaluate the performance of MAVRIC with the latest cross-section library in calculating the surface dose rates of a realistic spent-fuel storage cask. Detailed dose rate profiles over the cask side and top surfaces were calculated, and the results were compared with our previous work using SAS4 and MCNP. In order to duplicate the same source model, the MAVRIC code has been modified to accommodate a user-defined axial source distribution. The comparison among the three codes was evaluated in terms of their accuracies and computational efficiencies. For the gamma-ray sources, the MAVRIC-calculated results are more accurate than SAS4 and consistent with those predicted by the continuous-energy MCNP calculations. Meanwhile, its computational efficiencies are comparable to the performance of the TORT-coupled MCNP calculations. For the fuel neutron source, the MAVRIC calculation with broad-group cross sections cannot give satisfactory result, and its computational performance is also a factor of [approximately]10 less efficient than that of TORT-coupled MCNP. With a fine-group cross-section library, MAVRIC can provide a better prediction but still underestimates the surface dose rates of the cask by 15 to 30%.

 
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