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Nuclear Technology Volume 182, 2013 - Issue 2
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Regular Technical Paper

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

Y. F. ChenNational Tsing Hua University, Department of Engineering and System Science Hsinchu, Taiwan
,
Y. F. ChiouInstitute of Nuclear Energy Research, Taoyuan, Taiwan
,
S. J. ChangInstitute of Nuclear Energy Research, Taoyuan, 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
&
R. J. SheuNational Tsing Hua University, Department of Engineering and System Science Hsinchu, Taiwan;National Tsing Hua University, Institute of Nuclear Engineering and Science Hsinchu, Taiwan;National Synchrotron Radiation Research Center Hsinchu Science Park, Hsinchu, Taiwan
Pages 224-234 | Published online: 10 Aug 2017

REFERENCES

  • “Safety Analysis Report for the Independent Spent Fuel Storage Installation in Nuclear Power Plant 1,” Taiwan Power Company; http://www.aec.gov.tw/www/other/index_01_3_1.php (2007)(in Chinese)(current as of Jan. 18, 2012).
  • A. Haghighat and J. C. Wagner, “Monte Carlo Variance Reduction with Deterministic Importance Functions,” Prog. Nucl. Energy, 42, 25 (2003).
  • M. D. Dehart and O. W. Hermann, “An Extension of the Validation of the SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel,” ORNL/TM-13317, Oak Ridge National Laboratory (1996).
  • O. W. Hermann and M. D. Dehart, “Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel,” ORNL/TM-13315, Oak Ridge National Laboratory (1998).
  • B. L. Broadhead et al., “Evaluation of Shielding Analysis Methods in Spent-Fuel Cask Environments,” Nucl. Tech-nol., 117, 206 (1997).
  • H. Taniuchi and F. Matsuda, “Dose Rate Measurements and Calculation of TN-12/2 Packages,” Nucl. Technol., 127, 88 (1999).
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  • O.W. Hermann and C. V. Parks, “SAS2H:ACoupled One-Dimensional Depletion and Shielding Analysis Module,” NUREG/CR-0200, Rev. 6, Vol. I, Sec. S2, ORNL/NUREG/CSD-2/V2/R6, Oak Ridge National Laboratory (2000).
  • D. E. Peplow, “MAVRIC: MONACO with Automated Variance Reduction Using Importance Calculations,” ORNL/TM-2005/39, Version 6, Vol. I, Sec. S6, Oak Ridge National Laboratory (2009).
  • “SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations,” ORNL/TM-2005/39, Version 6, Vols. I, II, and III, Oak Ridge National Laboratory (2009).
  • D. E. Peplow, “Monte Carlo Shielding Analysis Capabilities with MAVRIC,” Nucl. Technol., 174, 289 (2011).
  • “Conversion Coefficients for Use in Radiological Protection Against External Radiation,” ICRP Publication 74, Ann. ICRP 26(3/4), International Commission on Radiological Protection (1996).
  • R. J. Sheu and S. H. Jiang, “Cosmic-Ray Induced Neutron Spectra and Effective Dose Rates near Air/Ground and Air/Water Interfaces in Taiwan,” Health Phys., 84, 92 (2003).
  • Y. F. Chen et al., “Surface Dose Rate Calculations of a Spent-Fuel Storage Cask by Using MAVRIC and Its Comparison with SAS4 and MCNP,” Nucl. Technol., 175, 343 (2011).
  • M. D. Dehart, “TRITON: A Two-Dimensional Transport and Depletion Module for Characterization of Spent Nuclear Fuel,” ORNL/TM-2005/39, Version 6, Vol. I, Sec. T1, Oak Ridge National Laboratory (2009).

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