Dose rate measurements and calculation of TN-12/2 packages

To clarify the effect of each assumption irt a shielding analysis of a spen t-fuel package to reduce the safety margin, the measured and calculated dos e rates around a package are compared. Neutron and gamma-ray dose rates,cer e 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 rent counter. Calc ulations were made using the SAS4M and MCNP codes based on detailed package and fuel assembly information, and the calculated and measured results wer e 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 e nds. 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 overes timate the measured ones br a factor of 2. In the bottom area, the discrepa ncy is within 40%. With respect to neutron dose rate, SAS4M and MCNP produc e 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 I m fr om th e surface; MCNP also overestimates, but the factor is lower. At the top, th e overestimation is much larger at the surface. At the bottom, there is goo d agreement. The causes of the differences between measurements and calculation using da ta front a safety analysis report are discussed. One Of the major reasons f or the difference is that the calculation model uses the minimum values req uired for thickness and density that were used in the safety analyses to ob tain conservative results. The angular dependence of the detector response and the effective center of the actual detector also affect the surface neu tron dose rate values obtained by measurement. lit addition, the burnup pro file 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 investigati on of the Co-60 source is necessary for future work.