Interim dry cask storage systems comprising AISI 304 or 316 stainless steel canisters have become critical for the storage of spent nuclear fuel from light water reactors in the Republic of Korea. However, the combination of microstructural sensitization, residual tensile stress, and corrosive environments can induce chloride-induced stress corrosion cracking (CISCC) for stainless steel canisters. Suppressing one or more of these three variables can effectively mitigate CISCC initiation or propagation. Surface-modification technologies, such as surface peening and burnishing, focus on relieving residual tensile stress by introducing compressive stress to near-surface regions of materials. Overlay coating methods such as cold spray can serve as a barrier between the environment and the canister, while also inducing compressive stress similar to surface peening. This approach can both mitigate CISCC initiation and facilitate CISCC repair. Surface-painting methods can also be used to isolate materials from external corrosive environments. However, environmental variables, such as relative humidity, composition of surface deposits, and pH can affect the CISCC behavior. Therefore, in addition to research on surface modification and coating technologies, site-specific environmental investigations of various nuclear power plants are required.

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

In the case of dry storage facilities, slipping of the cask or tip-over are dangerous phenomena. For this reason, in dry storage facilities, measures against slipping and tip-over or related safety evaluations are important. Accidental conditions that can cause cask slippage and tip-over in dry storage facilities include natural phenomena such as floods, tornadoes, tsunamis, typhoons, earthquakes, and artificial phenomena such as airplane crashes. However, among natural phenomena, earthquakes are the most important natural phenomenon that causes tip-over. Also, many people had the stereotype that Korea is an earthquake-safe zone before 2016. However, earthquakes become a major disaster in Korea due to the 2016 Gyeongju earthquake and the 2017 Pohang earthquake, followed by the Goesan earthquake in October 2022. In this paper, seismic analysis was performed based on dry storage facilities including multiple casks. Design variables for the construction of an analysis model for dry storage facilities were investigated, and seismic analysis was performed. To evaluate tip-over accident during earthquake, seismic load was used from 0.2 g PGA to 0.8 g PGA and these earthquakes were followed Design Response Spectrum (DRS) in RG 1.60. The friction coefficient of concrete pad was used from 0.2 to 1.0. As a result of the analysis, tip-over accident could not find in the analysis from 0.2 g to 0.6 g. However, tip-over was appeared at friction coefficients of 0.8 and 1.0 at 0.8 g PGA. Tip-over angular velocity of cask was derived by seismic analysis and was compared with formula and tip-over analysis results. As a result, a generalized dry storage facility analysis model was proposed, and dry storage facility safety evaluation was conducted through seismic analysis. Also, tip-over angular velocity was derived using seismic analysis for tip-over analysis.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties are needed to evaluate the structural integrity of the concrete structures. Especially, it has been known to be difficult to estimate the resulted damage precisely where an aircraft collides with the degraded concrete structures at a high strain rate. In this study, damage evaluation of concrete overpack due to aircraft collisions was conducted. First, in order to verify the impact analysis methodology, the aircraft impact analysis of plane concrete overpack was performed and compared with the test results previously conducted by our research team. Then, the impact analysis for the overpack of KORAD21C was performed. In the future, the radiation shielding analysis will be performed under the conditions to evaluate whether or not the radiation shielding ability is maintained.

In concrete structures exposed to chloride environments such as seashore structures, chloride ions penetrate into the concrete. Chlorine ions in concrete react with cement hydrates to form Friedel’s salt and change the microstructure. Changes in the microstructure of concrete affect the mechanical performance, and the effect varies depending on the concentration of chloride ions that have penetrated. However, research on the mechanical performance of concrete by chloride ion penetration is lacking. In this study, the effect of chloride ion penetration on the mechanical performance of dry cask concrete exposed to the marine environment was investigated. The mixture proportion of self-compacting concrete is used to produce concrete specimens. CaCl2 was used to add chlorine ions, and 0, 1, 2, and 4% of the binder in weight were added. To evaluate the mechanical performance of concrete, a compressive strength test, and a splitting tensile strength test were performed. The compressive strength test was conducted through displacement control to obtain a stress-strain curve, and the loading speed was set to 10 με/sec, which is the speed of the quasi-static level. The splitting tensile strength test was performed according to KS F 2423. As a result of the experiment, the compressive strength increased when the chloride ion concentration was 1%, and the compressive strength decreased when the chlorine ion concentration was 4%. The effect of the chloride ion concentration on the peak strain was not shown. In order to present a stress-strain curve model according to the chloride ion concentration, the existing concrete compressive stress-strain models were reviewed, and it was confirmed that the experimental results could be simulated through the Popovics model.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. During long-term storage, there are several mechanisms that could result in the degradation of spent nuclear fuels, and the temperature is the most important parameter to predict and estimate the degradation behaviors. Therefore, thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. Herein, Explicit CFD analysis model is introduced and validated by estimating the thermal hydraulic response of the dry storage system that is Dry Cask Simulator (DCS). Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to assess degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT is used to simulate DCS, and the test condition of 0.5 kW decay heat and 100 kPa helium pressure was investigated in this study. In case of peak cladding temperature (PCT), PCT from the experiment was 376 K while that of CFD was 374 K. It implies CFD simulation gives good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also show good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

Concrete structures of spent nuclear fuel interim storage facility should maintain their shielding ability and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain its unique functions (shielding and structural integrity). Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties of the concrete are required to evaluate the structural integrity of the concrete structures. Unlike the calculated damage results for the static deformation of the concrete structure, it is very difficult to accurately estimate the damage values of the degraded concrete structures where an aircraft collides at a high strain rate. Therefore, the present authors have a plan to establish a database of the dynamic material properties of deteriorated concrete and implement to a Finite Element Analysis model. Prior to that, dynamic increase factors described in a few technical specifications were investigated. The dynamic increase factor represents the ratio of the dynamic to static strength and is normally reported as function of strain rate. In ACI-349, only the strain rate is used as a variable in the empirical formula obtained from the test results of specified concrete strengths of 28 to 42 MPa. The maximum value of dynamic increase factor is limited to 1.25 in the axial direction and 1.10 in the shear direction. On the other hand, in the case of the CEB model, static strength is included as variables in addition to the strain rate, and a constitutive equation in which the slope changes from the strain rate of 30 /s is proposed. As plotting the two dynamic increase factor models, in the case of ACI, it is drawn as a single line, but in the case of CEB, it is plotted as multiple lines depending on the static strength. The test methods and specimen sizes of the previously performed tests, which measured the concrete dynamic properties, were also investigated. When the strain rate is less than 10 /s, hydraulic or drop hammer machines were generally used and the length of the specimens was more than twice the diameter in most cases. However, in the case of Split Hopkinson Pressure Bar tests, the small size specimens are preferred to minimize the inertia effect, so the specimens were small and the length was less than twice the diameter. We will construct the dynamic properties DB with our planned deteriorate concrete specimen test, and also include the dynamic property data already built in the previous studies.