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Technical Paper

Dose Evaluation for an Independent Spent-Fuel Storage Installation Using MAVRIC

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Pages 335-342 | Published online: 10 Apr 2017

References

  • “Safety Analysis Report for the Independent Spent Fuel Storage Installation in Nuclear Power Plant 1,” Taiwan Power Company, (2007) (in Chinese); http://www.aec.gov.tw/www/other/index_01_3_1.php (current as of Apr. 15, 2010).
  • J. N. WANG, C. H. LU, K. W LEE, U. T. LIN, and S. H. JIANG, “Study of the Site Dose Rate for the ISFSI Facility with Monte Carlo and Deterministic Methods,” Nucl. Technol., 168, 101 (2009).
  • 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).
  • “MCNP—A General Monte Carlo N-Particle Transport Code,” Version 4C, LA-13709-M, J. F. BRIESMEISTER, Ed., Los Alamos National Laboratory (2000).
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  • 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).
  • J. S. TANG and M. B. EMMETT, “SAS4: A Monte Carlo Cask Shielding Analysis Module Using an Automated Biasing Procedure,” ORNL/TM-2005/39, Version 5.1, Vol. I, Sec. S4, Oak Ridge National Laboratory (2006).

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