Year
2007
File Attachment
294.pdf219.29 KB
Abstract
The transport cask shielding analysis in Japan is carried out using the 2-dimensional Sn transport calculation code DOT2) and cross section set DLC-23/CASK3). Cross section set DLC-23/CASK was created for transport cask shielding calculation in 1973. Since this cross section set is user-friendly, it has been widely used in the shielding evaluation relevant to a spent fuel transport cask, and it has many track records. On the other hand, it was comparatively old aged, and has been pointed out that estimation accuracy is reduced in the specific condition such as particles penetrate in iron. The main reason why DLC-23/CASK has been long used as a standard cross section for transport cask in Japan is that we had no other more appropriate cross section data set for transport cask shielding calculation. For this reason, based on the newest evaluated nuclear data library in Japan, cross section data set SFCX-J33 for transport cask shielding calculation was developed. SFCX-J33 is the cross section set that improved the defects of DLC-23/CASK mentioned above. The main improvements are as follows. (1) The number of neutron energy groups was increased to 50 groups from 22 groups of DLC-23/CASK for the improvement in estimation accuracy. (2) The JENDL-3.3 developed by Japan Atomic Energy Agency was used as the evaluated nuclear data library. (3) SFCX-J33 was produced to be as user-friendly as DLC-23/CASK. Since estimation accuracy needed to be verified about developed cross section data set SFCX-J33, the shielding examination was performed using the main materials used for a transport cask. The shielding experiment was conducted at the shielding experiment institution of the National Maritime Research Institute in Japan, and neutron and secondary gamma ray measurements were performed using Cf-252 as the neutron source. Verification of SFCX-J33 was carried out by performing experiment benchmark calculation using the shielding experiment result