[en] The KN-12 transport cask is designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U)F package. W.H 14x14, 16x16 and 17x17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50000 MWD/MTU and minimum cooling time of 7 years being used in Korea are loaded and subsequently transported under dry and wet conditions. The containment boundary of the KN-12 cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the KN-12 cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the KN-12 cask are satisfied by maintaining a maximum air reference leak rate of 2.7x10^-4 ref cm³/sec or a helium leak rate of 3..3x10^-4 cm³/sec for normal transport conditions and for hypothetical accident conditions

[en] The spent fuel transport casks and spent fuel storage casks must be evaluated to dissipate the decay heat from spent fuel assemblies to the fuel basket and from the fuel basket to the outer cask surface. No active systems are required for removal and dissipation of the decay heat from spent fuel assemblies that is loaded within the casks. The fuel assemblies are very difficult to be modeled explicitly, i.e., fuel pellet, fuel cladding are not modeled separately on their own, but instead, they are available to be modeled as solids with homogeneous effective properties making no distinction between the different properties and heat transfer characteristics of cladding, pellet, spaces between rods, and gaps between pellet and cladding. This effective thermal property method will reduce analysis time and cost for thermal analysis of the cask. In this paper the effective thermal conductivity through a cross section of the fuel region of the fuel basket is calculated from a detailed two dimensional slice model of the traverse section of W.H 17x17 fuel assembly using FLUENT code based on the finite volume method. The effective thermal conductivity is found to model sufficiently the heat transfer by radiation and conduction between the fuel rods and between the fuel rods and the fuel basket in which the fuel assemblies reside, therefore this method could be applied to the thermal analyses of the transport casks and the storage casks

[en] We solve the simple physical model for liquid pool spreading with vaporization semi-analytically for the first time, using perturbation techniques. The results are compared with those obtained using numerical methods. We use the evaporation rate per unit area as a perturbation parameter, and first-order solutions are obtained for continuous and instantaneous release. The two solutions are nearly identical with respect to the pool radius. The pool volumes are nearly the same at the early stage of the spread and then start to diverge

[en] Since 1992, the commence of operation of concrete CANISTER the dry storage facility at Wolsong, there has never been a recorded incident or accident that might be able to lead to uncontrolled exposures to the workers and the public. The spent fuel dry storage facility has provided the Wolsong CANDU nuclear power station with an extra outdoor storage facility to store spent fuel in a safe, secure and reliable manner. Recently in Korea, a number of new dry storage system are under development and one of them is in the middle of licensing activities. However is may be safe, as for all activities involving radioactive sources, there is a potential risk associated with the normal operation of that facility as well as potential abnormal events involving fuel transfer and fuel storage. In this paper, a methodology to evaluate radiological consequences due to the dry storage system for Spent Fuel is introduced in associated with radiological principles and regulatory standards. Detailed evaluations will be performed based on the methodology presented in this paper for the module is implemented into the designated site

[en] A highly stable quartz crystal microbalance (QCM) sensor showing frequency stability and a very low noise level has been developed for measurements of water vapor. The long-term drift was < 0.05 Hz/h over a period of 10 h, and the short-term rms (root-mean-square) noise was < 0.015 Hz. The developed QCM sensor employing a poly(methyl methacrylate) (PMMA) polymer film as a hygroscopic layer on the front electrode of the quartz crystal was used to measure the water vapor concentration and showed good characteristics, good linearity, and fast response for the water vapor contents covered in this study, 175 ∼ 17,510 ppm_v. 2280 - 2329 Dr. Moon

[en] In this study, maximum exposure rate at DAW(Dry Active Waste) drum surface which is satisfying regulation limit was calculated for conceptual design of IP(Industrial Package). DAW can be classified as combustible and non-combustible waste and the calculation was conducted for single and mixed radionuclide for each type of waste. In case of combustible waste that mixed radionuclide is uniformly distributed, the maximum exposure rates at drum surface were 3.60E-01, 8.85E-01 and 1.27E+01 mSv/hr for IP Type 1, 2-a and 2-b, respectively. and 3.60E-01, 8.85E-01, 1.27E+01 mSv/hr for single radionuclide(Co-60). In case of non-combustible waste that mixed radionuclide is uniformly distributed, the maximum exposure rates at drum surface were 7.14E-01, 1.83E+00, 2.69E+01 mSv/hr for IP Type 1, 2-a and 2-b, respectively. and 7.13E-01, 1.81E-01, 2.62E+01 mSv/hr for single radionuclide(Co-60), Through this study, the maximum amount of DAW can be transported by IP was suggested as maximum exposure rate at drum surface and the calculation for the other types of waste will be conducted

[en] Two heat transfer tests were performed as a fabrication performance test to demonstrate the heat transfer capability of the KN-12 spent nuclear fuel transport cask. The tests were conducted under normal conditions of transport with a total heat load of 12.6kW to simulate the design heat load of the cask. The heat load was best represented by twelve electrical dummy heaters, which were designed to simulate actual configurations and conditions of twelve PWR spent nuclear fuel assemblies. The test determined steady state temperatures on the outer surfaces of the cask and impact limiters and within the fuel basket. The steady state temperatures were compared to the calculated temperatures to determine the accuracy of the design calculations. The intention of this paper is to evaluate test results which were measured during the heat transfer test for the KN-12 cask. The evaluation was done using maximum values for different cask components which were calculated for the Safety Analysis Report of the KN-12 transport cask. The test temperatures were described very well by the calculated maximum component temperatures and the calculated component temperatures were higher and therefore conservative

[en] KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr

[en] Large eddy simulation (LES) was performed to investigate backdraft in reduced-scale compartments using the fire dynamics simulator. Mixing-controlled fast chemistry combustion model was adopted in the LES predictions of the backdraft phenomenon. The effect of the opening geometry of the compartment on the critical fuel percentage required for backdraft occurrence was numerically investigated by considering compartments with a door or a slot opening. The LES with the mixing-controlled fast chemistry combustion model provided reasonable results for the gravity current and backdraft development process in the reduced-scale compartments. The predicted results of the critical fuel percentage for backdraft occurrence were in good agreement with those obtained by previous experiments on door and slot opening geometries. The LES also predicted the trend in peak pressure with increasing fuel concentration inside the compartment reasonably well. The oxygen concentration of the entrained airflow inside the compartment affected the peak pressure during backdraft development. The peak pressures for the compartment with a slot opening were higher than those for a compartment with a door opening at the same fuel concentration. The difference in the peak pressures between the two compartment geometries was attributed to the difference in the entrained oxygen concentration caused by different ignition times.

[en] We evaluated the importance of redistribution and 24 hour reinjection images in Tl-201 SPECT assessment of myocardial viability after acute myocardial infarction (AMI). We performed dipyridamole stress-4 hour redistribution-24 hour reinjection Tl-201 SPECT in 43 patients with recent AMI (4-16 days). The myocardium was divided into 16 segments and perfusion grade was measured visually with 4 point score from 0 to 3 (absent uptake to normal uptake). A perfusion defect with stress score 2 was considered moderate. A defect was considered severe if the stress score was 0 or 1 (absent uptake or severe perfusion decrease). Moderate defect on stress image were considered viable and and segments with severe defect were considered viable if they showed improvement of 1 score or more on redistribution or reinjection images. We compared the results of viability assessment in stress-redistribution and stress-reinjection images. On visual analysis, 344 of 688 segments (50%) had abnormal perfusion. Fify two (15%) had moderate perfusion defects and 292 (85%) had severe perfusion defects on stress image. Of 292 severe stress defects, 53 were irreversible on redistribution and reversible on reinjection images, and 15 were reverseble on redistribution and irreversible on reinjection images. Two hundred twenty four of 292 segments (76.7%) showed concordant results on stress-redistribution and stress- reinjection images. Therefore 24 hour reinjection image changed viability status from necrotic to viable in 53 segments of 292 severe stress defect (18%). However, myocardial viability was underestimated in only 5% (15/292) of severe defects by 24 hour reinjection. The 24 hour reinjection imaging is useful in the assessment of myocardial viability. It is more sensitive than 4 hour redistribution imaging. However, both redistribution and reinjection images are needed since they complement each other