To clarify the effect of each assumption in a shielding analysis of a spent-fuel package to reduce the safety margin, the measured and calculated dose rates around a package are compared. Neutron and gamma-ray dose rates were measured at many points around a TN-12/2 transport package loaded with 1.5-yr-cooled spent fuel using an ionization chamber and a rem counter. Calculations were made using the SAS4M and MCNP codes based on detailed package and fuel assembly information, and the calculated and measured results were then compared. For the sides of the package, the discrepancy between the measured and calculated gamma-ray dose rates is within 50% except at both ends. There are discrepancies of a factor of 2 or 3 in the results for both end surfaces. In the top region, the calculated gamma-ray dose rates overestimate the measured ones by a factor of 2. In the bottom area, the discrepancy is within 40%. With respect to neutron dose rate, SAS4M and MCNP produce different results. On the sides, the SAS4M calculation overestimates the measured dose rates by a factor of 2 at the surface and 1.7 at 1 m from the surface; MCNP also overestimates, but the factor is lower. At the top, the overestimation is much larger at the surface. At the bottom, there is good agreement.

The causes of the differences between measurements and calculation using data from a safety analysis report are discussed. One of the major reasons for the difference is that the calculation model uses the minimum values required for thickness and density that were used in the safety analyses to obtain conservative results. The angular dependence of the detector response and the effective center of the actual detector also affect the surface neutron dose rate values obtained by measurement. In addition, the burnup profile of the spent fuels affects not only the neutron dose rate but also the gamma-ray dose rate at both ends of a package. A more detailed investigation of the 60Co source is necessary for future work.