Measurement of dose-equivalent rates around a cask and Monte Carlo analysis with actual configuration of fuel basket

Gamma ray and neutron dose-equivalent rate distributions are measured with an ionization-type gamma-ray survey meter and a moderator-type neutron survey meter respectively around a TN-12A spent fuel transport cask, and the measured dose-equivalent rate distributions are analyzed by the Monte Carlo method. Two models for the aluminum-alloy fuel basket are considered in the Monte Carlo code MCNP 4B. In one case the configuration of the basket with 12 spent fuel assemblies is modeled in detail and the other is the homogenized basket as employed in the Sn code DOT 3.5. In addition, the burn-up distribution is taken into account to generate source neutrons and gamma rays in the z-axis of the spent fuel assemblies in both cases. The essential difference in the dose-equivalent rates is obtained from the Monte Carlo calculations employing the homogenized model and the actual configuration of the basket. Due to employing the actual configuration, the gamma ray and neutron dose-equivalent rates reduce to 67% and 80%, respectively as compared with the homogenized model on the surface of the TN-12A cask. As the results, the good C/Es are obtained: for the neutron it is approximately 1.0 and 1.25 for the gamma ray it is at the center of the cask surface, respectively. The effect of the burn-up distribution appears clearly at the off-center cask surface, and in particular, the neutron dose-equivalent rates come close to the measured ones. (author)