In this study, lab-scale vacuum drying experiments were conducted using canisters with a volume of 196 L and a height of 1 m, along with vacuum pumps with various intake capacities of 100, 200, 400, and 600 L/min. Through these experiments, when performing direct pressure-reduction vacuum drying tests, we calculated the evaporation rate of residual water over time. We identified a significant increase in the evaporation rate near approximately 40-50 torr, marking a critical transition point where residual water undergoes a phase change from liquid to vapor. This zone was experimentally determined and subsequently termed the "Evaporation-Acceleration Zone." Before entering this zone, the likelihood of ice formation is theoretically negligible even when employing high-capacity pumps. Based on these observations, our research team concluded that direct pressure reduction is feasible and opted for a strategy to maximize evaporation efficiency by prolonging the duration within the phase change boundary, rather than employing the traditional stepwise pressure reduction method. The conventional stepwise pressure reduction technique, widely used in the industry, exhibited an average evaporation efficiency of 9.66 mg/s. In contrast, when hold times at the phase change boundary were extended to 25 and 50 min, evaporation efficiencies were recorded at 18.526 mg/s and 12.698 mg/s, respectively, showing an increase of 1.92-fold and 1.31-fold. Additionally, initiating stepwise pressure reduction at 5 min into the phase change boundary led to a 1.62-fold improvement in efficiency. Beaker and simulated drainage hole tests yielded similar results, with evaporation rates measuring 9.668 mg/s and 9.256 mg/s, respectively.

This study presents a radiation-induced thermal conductivity degradation (TCD) model of zirconium as compared to the conventional UO2 TCD model. We derived the governing factors of the radiation-induced TCD model, such as maximum TCD value and temperature range of TCD. The maximum TCD value was derived by two methods, in which 1) experimental result of 32 % TCD was directly utilized as the maximum TCD value and 2) a theoretical approach based on dislocation was applied to derive the maximum TCD value. Further, the temperature range of TCD was determined to be 437–837 K by 1) experimental results of post-annealing of irradiation hardening as compared to 2) the rate theory and thermal equilibrium. Consequently, the radiation-induced TCD model of zirconium was derived to be fr=1−0.321+exp⁡{(T−637)/45} . Because the thermal conductivity of zirconium is one of the factors determining the storage and transport system, this newly proposed model could improve the safety analysis of spent fuel storage systems.

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. To safely store spent nuclear fuel in a dry storage system, residual water must be removed from the dry storage system canisters. The drying methods currently used to remove residual water from dry storage system canisters are vacuum drying and forced gas (helium) dehydration. It is known that ice can form when residual water in a canister is removed using vacuum drying. Therefore, a vacuum drying system and a small canister were designed, fabricated, and used to conduct vacuum drying tests to investigate whether the drying of residual water inside the canister was adequate. The formation of ice inside the canister during vacuum drying was also investigated. In the vacuum drying, the ice formation was found to be related to the surface area of water, and the wider the surface area of water, the faster the drying process. When all the residual water inside the canister is dried, the internal temperature of the canister did not change, and the pressure decreased sharply after reaching 3 torr. Drying was more done when the hold point was kept at a higher pressure as well as hold time was kept longer. Therefore, the hold time at each step and the hold point of pressure should be well set to prevent the ice formation. Vacuum drying method that reduces the pressure step by step is more desirable than a vacuum drying method using a vacuum pump having a slower flow rate. (C) 2021 Elsevier Ltd. All rights reserved.

The weight of spent nuclear fuel concrete storage casks usually exceeds 100 tons. Therefore, thermal testing of prototype casks is expensive and time consuming. In this study, thermal testing method using a scaled-down model of a storage cask was developed to perform the thermal testing of large-scale storage casks efficiently. Scaling analyses were performed to derive scale ratios between a prototype cask and the scaled-down model. Additionally, thermal analyses were performed on the prototype cask and half-scale model using the obtained scale ratios. The analysis results showed that scale ratios between the prototype cask and half-scale model were conserved well. Thermal tests were performed on the prototype cask and half-scale model. The analysis and test results were in good agreement. The thermal test results confirmed the thermal-fluid flow similarity between the prototype cask and half-scale model. (C) 2020 Elsevier Ltd. All rights reserved.

KORAD-B/II shipping packages are used to transport C4 concrete packages that are temporarily stored at the HANUL nuclear power plant. These packages must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standards No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify a KORAD-B/II shipping package as a Type B package, and state that this type of package must be able to withstand a temperature of 800 °C for a period of 30 min. It is desirable to conduct a test using a full-scale model of a shipping package when performing tests to evaluate its integrity. However, it is costly to perform a test using a full-scale model. Therefore, to evaluate the thermal integrity of a KORAD-B/II shipping package, thermal tests were conducted using a slice model. For comparison purposes, a thermal test was also carried out using a half-scale model. In the first thermal test using a slice model, the maximum surface temperature of the cask body was higher than the permitted maximum temperature limits owing to incomplete combustion. In the second thermal test using a slice model and in the thermal test using a half-scale model, the maximum temperature of the cask body was lower than the permitted maximum temperature limit. Therefore, the thermal integrity of the KORAD-B/II shipping package could be considered to be maintained. The temperature results from the thermal test using a slice model were higher than those of the thermal test using a half-scale model. Therefore, the effect of flame on a single-layer shipping package without neutron shielding, such as the KORAD-B/II shipping package, seems to be affected by the reduction in the time rather than the size reduction.

The flow characteristics inside a fuel basket containing a spent nuclear fuel assembly were analyzed in this study. The geometry of a nuclear fuel assembly is very complicated; hence, a porous media model and an effective thermal conductivity were used to simplify the fuel assembly geometry. The permeability was calculated in three different ways, namely, a theoretical approach using the friction factor assuming laminar flow, a computational fluid dynamics (CFD) calculation using shear stress, and a CFD calculation using pressure drop. The results of the CFD calculations by shear stress and pressure drop showed a good agreement. The result of the theoretical approach was 30% less than the CFD results because the theoretical approach did not fully consider the flow resistance resulting from the complex geometry of the spacer grid and the fuel rods. The calculations were repeated using simplified models based on the repeated geometry in the system. The simplified model with periodic boundary conditions was used to accurately calculate the permeability while increasing the CFD calculation efficiency. (C) 2018 Elsevier Ltd. All rights reserved.

KORAD-B/II shipping packages, developed by the Korea Radioactive Waste Agency (KORAD), are used to transport C4 concrete packages, which are temporarily stored in the HANUL nuclear power plant. These packages must therefore meet the requirements of the Korea Nuclear Safety Security Commission Act 2017-56, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. The regulatory guidelines classify a KORAD-B/II shipping package as a Type B package, and state that this type of package must be able to withstand a temperature of 800 degrees C for a period of 30 min. It is desirable to use a full-scale model of a shipping package when performing tests to evaluate its integrity. However, performing a test using a full-scale model is costly. Therefore, to evaluate the thermal integrity of a KORAD-B/II shipping package, thermal tests were conducted using a slice model. For comparison, a thermal test was also carried out using a half-scale model. In the first thermal test using the slice model, the maximum surface temperature of the cask body was higher than the permitted maximum temperature limit because of incomplete combustion. In the second thermal test using the slice model and in the thermal test using the half-scale model, the maximum temperature of the cask body was lower than the permitted maximum temperature limit, thus proving that the thermal integrity of the KORAD-B/II shipping package could be maintained. The temperature results of the thermal test using the slice model were higher than those of the thermal test using the half-scale model. The effect of flame on a single-layer shipping package without neutron shielding, such as the KORAD-B/II shipping package, appears to be influenced more by a reduction in time rather than in size.

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask developed for this purpose should be able to adequately emit the decay heat from the spent nuclear fuel; it should also maintain the temperatures of the spent nuclear fuel assemblies within allowable values under normal and off-normal conditions and during an accident. However, since the thermal conductivity of concrete is low and the allowable temperature of concrete is lower than that of steel, the concrete storage cask must be designed to have heat-removal capability with appropriate reliability. A passive heat-removal system was designed to maintain the temperatures of the fuel assembly cladding material and concrete storage cask components within the allowable limits; it consists of four air inlets and four air outlets with openings that are covered by mesh screens to prevent debris or wildlife from entering the ventilation ducts. Thermal tests were performed to evaluate the effect of the mesh size of each screen on heat-removal performance of the concrete storage cask. The screen mesh size was estimated to have an insignificant effect on the temperature rise of the canister surface and the over-pack surface, but it had a considerable effect on the temperature rise of the components of the over-pack body. As the screen mesh size decreased, the heat-removal by natural convection cooling through the passive heat-removal system was reduced, and the temperature of the concrete storage cask rose. (C) 2017 Elsevier Ltd. All rights reserved.

The storage and transport system are being developed in Korea. The spent fuels stored temporarily in wet storage at reactor site are scheduled to be transported to the interim storage in near term in Korea. To prepare the spent fuel transportation and storage, the dual purpose cask and concrete cask have been developed and submitted for the license approval.Safety evaluation test was conducted with 1/3 scaled model and so was analysis. Since numerical simulation results can be different based on the analysis models, the reliable FE model was generated through several revising process. The comparison between simulation and test was conducted by numerical simulation using this FE model.The reliable FE model made it possible to be proceeded stably by analysis and generated more realistic results compared to those of test. By the comparison between analysis and test results, the validity and appropriateness of the modeling technique and analysis methodology were verified.

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask developed for this purpose should be able to adequately emit the decay heat from the spent nuclear fuel. Moreover, the concrete storage cask must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values under normal and offnormal conditions and during an accident. However, the thermal conductivity of concrete is low and the allowable temperature of concrete is lower than that of steel. Therefore, the concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. Accordingly, a passive heat removal system was designed to maintain the temperatures of the fuel assembly cladding material and concrete storage cask components within the allowable limits. The passive heat removal system consists of four air inlets and four air outlets with openings that are covered by mesh screens to prevent debris or wildlife from entering the ventilation ducts. The effect of each screen on the heat removal of the concrete storage cask depends on its mesh size. Therefore, thermal tests were performed to evaluate the heat removal performance in accordance with the mesh size of the screen. The screen mesh size was estimated to have an insignificant effect on the temperature rise of the canister surface and the over-pack surface, but it had a considerable effect on the temperature rise of the components of the over-pack body. As the screen mesh size decreased, the heat removal by the natural convection cooling through the passive heat-removal system was reduced, and the temperature of the concrete storage cask rose.

Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 degrees C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 degrees C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the body of the test model through the heat transfer fin. The maximum temperatures of the neutron shielding at the part where the heat transfer fin was installed were 155 degrees C. However, those in the part where the heat transfer fin was not installed were 183 degrees C. The neutron shielding was therefore adequately protected by the heat transfer fin. (C) 2016 Elsevier B.V. All rights reserved.