A comparison of dose rate calculations for a spent fuel storage cask by using MCNP and SAS4-Reference-Cited by-同舟云学术

A comparison of dose rate calculations for a spent fuel storage cask by using MCNP and SAS4

Published:2008-12 Issue:12 Volume:35 Page:2296-2305
ISSN:0306-4549
Container-title:Annals of Nuclear Energy
language:en
Short-container-title:Annals of Nuclear Energy

Author:

Chen A.Y.,Chen Y.F.,Wang J.N.,Sheu R.J.,Liu Y.-W.H.,Jiang S.H.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference20 articles.

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2. Bowman, S.M., Dunn, M.E., Hollenbach, D.F., Jordan, W.C., 2006. SCALE Cross-section Libraries. ORNL/TM-2005/39, Version 5.1, vol. III, Sect. M4. Oak Ridge National Laboratory, Oak Ridge, TN.

3. Briesmeister, J.F. (Ed.), 2000. MCNP – A General Monte Carlo N-Particle Transport Code, Version 4C, LA-13709-M. Los Alamos National Laboratory, USA.

4. Evaluation of shielding analysis methods in spent-fuel cask environments;Broadhead;Nucl. Tech.,1997

5. Effective biasing schemes for duct streaming problems;Broadhead;Radiat. Prot. Dosim.,2005

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