SHIELDING INTEGRITY TEST FOR PRE-SERVICE INSPECTION OF KN-12 TRANSPORT CASK

The calculated and measured shielding evaluation of the cask for 12 spent nuclear fuels has been performed to confirm the shielding integrity of cask for pre-service inspection of KN-12 transport cask. The spent fuel source terms were calculated by Origen-Arp and Shielding calculations were made with MCNP5. The shielding evaluation was performed for the cask with 12 particular fuel assemblies and a uniformly distributed Co-60 gamma source was assumed for the plenum, top nozzle, and bottom nozzle. All measurements were carried out with an isolated cask in spent fuel decontamination pit at KORI Nuclear Power Plant in Korea and discrete measurements were performed at the 0 o axis, 90 o, 180 o, and 270 o for getting dose rate profiles. Neutron dose rate measurements were performed with the commercial device PTS200 from Atlan Tech and the gammas dose rates were measured with the SURVEYOR 2000 from BICRON. By comparison of calculated neutron dose rates with measured dose rates profiles for the 0 o axis, 90 o, 180 o, and 270 o at axial cask surface, the C/E ratio of the axial neutron dose rates is around 1.0 ~ 2.56 except the trunnion region. Dose rates calculated for the trunnion region show the conservative results, because the trunnions are not modeled in this model. Several points at top lid and bottom lid are also compared between calculated and measured values. It is shown that the C/E ratios at top lid region are some higher than others. Gamma dose rates were also compared each other with same procedure of neutron case. The C/E ratio at center of the cask side is 3.0 at the 0 o axis, 90 o, 270 o and 2.3 at the 180 o. The result shows that C/E ratio is 2.1 ~ 3.4 at bottom regions, and 1.0 ~ 1.9 at top of the cask. The comparison between the calculated and measured dose rate shows good agreement for all axial, top and bottom profiles, i.e. the same principle trends with an overestimation of the measured dose rate by the calculations.