Accuracy verification study of the new cross section library set, SFCX-J33', by comparison of calculation using two-dimensional discrete ordinate code and experiment using 252Cf fission source

Discrete ordinate calculation codes and multi-group cross section library sets are mostly used for shielding analysis of transport casks. The DLC-23/CASK cross section library set which is developed at ORNL in 1973 is widely used. But this library set includes old data. Therefore, the accuracy of calculation deteriorates in some specific cases, for example, when neutrons penetrate into iron plates. To solve these problems, SFCX-J33, a new cross section library set for spent fuel transport cask, is developed based on JENDL3.3 which is the newest nuclear cross section data set in Japan. In this research, at first, neutron and photon ambient dose equivalents and neutron spectra were measured in the slab geometry composed of 252Cf neutron source pit and plates of stainless-steel, lead, resin and polyethylene. Secondly ambient dose equivalents were measured in the cask geometry composed of carbon-steel cylindrical vessel surrounded by a resin neutron shield encasing 252Cf neutron source. To compare with the measured data, the ambient dose equivalents and neutron spectra were calculated by a two-dimensional discrete ordinate code, DOT with the new cross section library set, SFCX-J33 under the slab and cask geometries. The ambient dose equivalent and neutron spectra calculated by DOT with SFCX-J33 were compared with those measured in the experiments using 252Cf spontaneous fission source and both results agreed well. It is shown that SFCX-J33 and DOT code are applicable to shielding analysis of transport cask. (author)