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Nuclear Technology Volume 175, 2011 - Issue 1: Special Issue on The 16th Biennial Topical Meeting of The Radiation Protection and Shielding Division
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Technical Paper

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

Y. F. ChenNational Tsing Hua University, Department of Engineering and System Science Hsinchu, Taiwan
,
R. J. SheuNational Synchrotron Radiation Research Center, Hsinchu Science Park Hsinchu, Taiwan
,
S. H. JiangNational Tsing Hua University, Department of Engineering and System Science Hsinchu, Taiwan;National Tsing Hua University, Institute of Nuclear Engineering and Science Hsinchu, Taiwan
,
J. N. WangInstitute of Nuclear Energy Research, Taoyuan 325, Taiwan
&
U. T. LinInstitute of Nuclear Energy Research, Taoyuan 325, Taiwan
Pages 343-350 | Published online: 10 Apr 2017

References

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  • “MCNP—A General Monte Carlo N-Particle Transport Code,” Version 4C, LA-13709-M, J. F. BRIESMEISTER, Ed., Los Alamos National Laboratory (2000).
  • A. Y. CHEN, Y. F. CHEN, J. N. WANG, R. J. SHEU, Y.-W. H. LIU, and S. H. JIANG, “A Comparison of Dose Rate Calculations for a Spent Fuel Storage Cask by Using MCNP and SAS4,” Ann. Nucl. Energy, 35, 2296 (2008).
  • “Safety Analysis Report for the Independent Spent Fuel Storage Installation in Nuclear Power Plant 1,” Taiwan Power Company (2007).
  • “Neutron and Gamma-Ray Fluence-to-Dose Factors,” ANSI/ANS-6.1.1-1991, American Nuclear Society, La Grange Park, Illinois (1991).

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