STUDY OF ANALYSIS METHODS OF SHIELDING CALCULATION CODES FOR CASKS

Year
2016
Author(s)
Ai Saito - Transnuclear, Tokyo 1-18-16, Shinbashi, Minato-ku, Tokyo 105-0004 JAPAN
Akihiko Terada - Transnuclear, Tokyo 1-18-16, Shinbashi, Minato-ku, Tokyo 105-0004 JAPAN
Dai Yokoe - Transnuclear, Ltd.
Hiroki Sakamoto - Transnuclear Tokyo 1-18-16, Shinbashi, Minato-ku, Tokyo 105-0004 JAPAN
Hiroaki Taniuchi - Transnuclear, Ltd.
File Attachment
F3033.pdf629.34 KB
Abstract
It is common in Japan to apply QAD-CGGP2R, DOT3.5 and DORT to safety analysis for transport and storage cask, and MCNP code is being introduced recently.In order to study the features of each analysis code, using a cask shielding benchmark problem, investigations are carried out in this paper concerning the following subjects.a) Influence of difference of nuclear data library on calculation results (DORT).b) Influence of number of angular quadratures on ray effect (DORT).c) Influence of homogenization of fuel region on calculation results (QAD-CGGP2R, MCNP)d) Influence of difference of analysis code on calculation results (QAD-CGGP2R, DORT, MCNP)e) Effectiveness of the latest method for variance reduction parameter setting (ADVANTG + MCNP)The results of the above investigations show features of each analysis code and that attentions shall be paid to setting of parameters of the analysis code when utilized. In addition, the latest knowledge concerning effectiveness of ADVANTG recently released from by ORNL (Oak Ridge National Laboratory) for variance reduction parameter setting of MCNP has been obtained.