A comparison and validation between the analysis and vibration test data of a nuclear fuel assembly were conducted. During the comparison and validation process, various parameters that govern the vibration behavior of the fuel assembly were determined, including nuclear fuel rod’s stiffness, spring constants of the dimple and spring of support structures, and damping coefficients. The calibration of the vibration analysis model aimed to find analysis parameters that can accurately simulate the vibration behavior of the test data. For calibration, power spectral density (PSD) diagrams were generated for both the measured signals from the test and the calculated signals from the analysis. The correlation coefficient between these two PSD plots was calculated. To find the analysis parameters, each parameter was defined as a variable with an appropriate range. Latin hypercube sampling was used to generate multiple sample points in the variable space. Analysis was performed for the generated sample points, and PSD plot correlation coefficients were calculated. Using the generated sample points and their corresponding results, a Gaussian Process Regression model was implemented for PSD plot correlation coefficients and the maximum PSD value. Based on the constructed surrogate model, the optimal analysis parameters were easily found without additional computations. Through this method, it was confirmed that the analysis model using the optimal parametes appropriately simulates the vibration behavior of the test.

In this study, a fracture evaluation of the spent nuclear fuel storage canister was conducted. Stainless steel alloys are typically used as the material for canisters, and therefore, a separate destructive evaluation is not required for safety analysis reports. However, in this research, a methodology for conducting a destructive evaluation was proposed for assessing the acceptability of cracks detected during in-service inspections for long-term storage due to reasons such as stress corrosion cracking. For the fracture evaluation, analytical equations provided in the design code such ASME were employed, and finite element method (FEM) based linear elastic fracture mechanics (LEFM) was performed to validate the effectiveness of the analytical equations. Impact analyses such as tip-over of the storage cask on a concrete pad were performed, and the fracture evaluation using stresses resulting from the impact analysis under accident conditions and residual stresses from welds were carried out. Through this research, geometric dimensions for cracks exceeding the fracture criteria were established.

It is very important that the confinement of a spent fuel storage systems is maintained because if the confinement is damaged, the gaseous radioactive material inside the storage cask can leak out and have a radiological impact on the surrounding public. For this reason, leakage rate tests using helium are required for certificate of compliance (CoC) and fabrication inspections of spent fuel storage cask. For transport cask, the allowable leakage rate can be calculated according to the standardized scenario presented by the IAEA. However, for storage cask, the allowable leakage rate is determined by the canister, facility, and site specific information, so it is difficult to establish a standardized leakage rate criterion. Therefore, this study aims to establish a system that can derive system-specific leakage test criteria that can be used for leakage test of actual storage systems. First, the variables that can affect the allowable leakage rate for normal and accident conditions were derived. Unlike transportation systems, for storage systems, the dose from the shielding analysis and the dose from the confinement analysis are summed up to determine whether the dose standard is satisfied, and even the dose from the existing nuclear facilities is summed up during normal operation condition. For this reason, the target dose is used as an input variable when calculating the allowable leakage rate for the storage system. In addition, the main variables are the distance from the boundary of the exclusive area, the number of cask, the inventory of nuclide material in the cask, the free volume, and the internal and external pressure. Utilizing domestic and US NRC guidelines, we derived basic recommended values for the selected variables. The GASPARII computer code that can evaluate the dose to the public under normal operating conditions was utilized. Using the above variables, the allowable leakage rate is calculated and converted to the allowable criteria for helium leakage rate test. The developed system was used to calculate the allowable leakage rate for normal and accident conditions for a hypothetical storage system. The leakage rate criteria calculation system developed in this study can be useful for CoC and fabrication inspections of storage systems in the future, and a GUI-based program will be built for user convenience.

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 this study, dynamic material testing was conducted on concrete cores extracted from HANARO exterior wall during seismic reinforcement construction. HANARO was constructed at the Korea Atomic Energy Research Institute in 1995, following strict nuclear quality assurance standards. In order to conduct the dynamic material testing of the extracted concrete cores, self-disposal had to be performed because the concrete cores were extracted and stored in a radiation controlled area. A self-disposal application was prepared and submitted based on the radionuclide analysis results, and it was finally approved in April 2023. Then, a test was performed by processing a specimen for dynamic property testing using a self-disposed concrete core. The concrete cores were processed to create specimens for dynamic material testing and the dynamic material testing was performed to obtain stress-strain diagrams according to the strain rate.

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.

Since the time to consider when evaluating leakage of spent fuel dry storage systems is very long, assumptions that continue to leak at the initial leakage rate are too conservative. Therefore, this study developed a dynamic methodology to calculate the change in leakage rate using time-varying variables and apply it to calculate the amount of radioactive leakage during the evaluation period. The developed dynamic methodology was then applied to calculate the leakage radiation source term for a hypothetical dry storage system and used to perform a public dose assessment. When applying the developed dynamic leakage rate evaluation methodology for more accurate confinement evaluation in case of containment damage of dry storage system, it was found that the change of leak rate with time is very insignificant if the hole diameter is small enough, and the leak rate decreases rapidly with time when a hole with a certain diameter or larger occurs. In the case of the accident condition, except when the hole is very large, it corresponds to the chocked flow condition, and the leak rate decreases rapidly as soon as the internal pressure is sufficiently lowered to enter the molecular and continuum flow region. In the case of a small hole diameter, the leakage volume is very small, so even if the dynamic methodology is applied, the evaluation results are not different from the case where the initial leakage rate continues, and when the hole diameter exceeds a certain value, the internal pressure drops according to the leakage volume, and the leakage rate decreases significantly. As a result of evaluating the dose to residents by applying the calculated radiation source term, it was confirmed that the dose criteria was exceeded when a hole with a diameter of about 4 μm occurred under off-normal conditions, and the dose standard was exceeded under accident conditions when a chocked flow occurred between the diameter of the hole and 2-3 μm, resulting in a rapid increase in the dose. The results of this study are expected to contribute to a more accurate evaluation of the confinement performance of storage systems, which will contribute to the design of optimal dry storage systems.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, building interim storage facilities or permanent disposal facilities should be considered. These storage facilities are divided into wet storage facilities and dry storage facilities. Wet storage facility is a method of storing SNF in SFP to cool decay heat and shielding radiation, and dry storage facility is a method of storing SNF in a cask and placing on the ground or storage building. However, wet storage facilities have disadvantages in that operating costs are higher than that of dry storage facilities, and additional capacity expansion is difficult. Dry storage facilities have relatively low operating costs and are relatively easy to increase capacity when additional SNFs need to be stored. For this reason, since the 1990s, the number of cases of applying dry storage facilities has been increasing even abroad. Dry storage facilities are divided into indoor storage facilities and outdoor storage facilities, and outdoor storage facilities are mostly used to take advantage of dry storage facilities. In the case of outdoor storage facilities, the cask in which SNFs are stored is placed on a designed concrete pad. During this storage, the boring heat generated by SNFs cools into natural convection and the cask shields the radiation that SNFs generates. However, if an accident such as an earthquake occurs and the cask overturns during storage, there may be a risk of radiation leakage. Such a tip-over accident may be caused by the cask slipping due to the vibration of an earthquake, or by not supporting the cask properly due to a problem in the concrete pad. Therefore, in the case of outdoor dry storage facilities, it is necessary to evaluate the seismic safety of concrete pads. In this paper, various soil conditions were applied in the seismic analysis. Soil conditions were classified according to the shear wave velocity, and the shear wave velocity was classified according to the ground classification criteria according to the general seismic design (KDS 17 10 00). The concrete pad was designed with a size that 8 casks can be arranged at regular intervals, and 11# reinforcing bars were used for the design of the internal reinforcement of the concrete pad according to literature research. The cask was designed as a rigid body to shorten the analysis time. The soil to which the elastic model was applied was designed under the concrete pad, and infinite elements were applied to the sides and bottom of the soil. The effect on the concrete pad and cask by applying a seismic wave conforming to RG 1.60 to the bottom of the soil was analyzed with a finite element model.

In order to construct and operate the dry storage systems, it is essential to confirm the safety of the systems through safety analysis. If the dry storage cask is damaged due to an accident, a large amount of radioactive material may be leaked to the outside and cause radiation exposure to surrounding workers and nearby public, so the effect thereof should be evaluated. Many input parameter are required in the confinement evaluation for accident condition, and in this study, the change in the confinement evaluation result according to the change of major input parameter is to be studied. In this study, we selected fractions of radioactive materials available for release from spent fuel, cooling time, and distance to exclusive area boundary as the major input parameter. In general, the release fraction suggested by NUREG-1536 has been used, but NUREG-2224 provides the fraction for high burn-up spent fuel in fire and impact accident conditions, unlike NUREG-1536 which provide a single value. In the case of the distance to exclusive area boundary, 100 to 800 m was considered, and in the case of the cooling time, 10 to 50 years was considered in this study. In order to compare the dose change by the parameter, we set up the hypothetical storage system. A storage cask of the system contain 21 PWR spent fuel assemblies with an initial enrichment of 4.5wt%, burnup of 45,000 MWD/MTU. During the accident condition, it is assumed that the cask is leaked at 1.0×10−7cm3·sec−1. Since the main dose criterion for accident conditions is 50 mSv of effective dose, effective doses are calculated in this study. In an accident condition, transuranic particulate contribute most of the doses, so the doses are determined according to the fraction for the particulate. Therefore, it was confirmed that the dose was almost the same as the fraction for the accident conditions in NUREG-1536 and the fraction for the impact accident conditions in NUREG-2224 is 3×10−5, but the dose was also 100 times higher as the fraction for the fire accident conditions in NUREG-2224 is 3×10−3. In the case of the cooling time, it was confirmed that the dose change according to the cooling time was not significant because the dose contribution of transuranic elements having very long half-life was very large. In the case of the distance, it was confirmed that the dose decreased exponentially as the atmospheric dispersion factor decreased exponentially with the distance.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, transporting the SNF from SFP to interim storage facilities or permanent disposal facilities should be considered. There are two ways to transport the SNF from a site to other site, one is the land transportation with truck or train, and the other is the maritime transportation with ship. The maritime transportation has some advantages compared with the land transportation. The maritime transportation method uses safer route which is far from populated area than land transportation method, and transport more weight than land transportation method. However, the cask should be loaded into the ship for the maritime transportation, and there is a possibility of a drop accident of the cask onto the ship. Therefore, it is necessary to evaluate the structural integrity of the cask and ship for the drop accident during the loading process. To evaluate the structural integrity of the cask and ship, it is necessary to determine the analysis conditions that caused the greatest damage in the drop accident. There may be various conditions such as the drop angle of the cask, the initial falling speed, the drop position onto the ship, the size of the ship, etc. This study set the drop angle of the cask and the drop position onto the ship as the simulation variables, which have high possibility to occur during cask drop. However, the others are excluded since they are controllable by worker. In this paper, various drop angle (0, 15, 30, 45, and 70 degree) of the cask were simulated to define the greatest damage condition. KORAD-21 cask model was used for Finite Element Analysis (FEA), and FEA was performed to simulate a horizontal drop (1 m drop). The strain-hardening material properties for the deck were used as HT36 steel. The Cowper-Symonds constitutive model for HT36 was used to consider the strain rate effect. A Tie-down structure for supporting the cask was modeled with the cask model which contained inner structures like canister, basket, etc. Structural integrity of the cask and tie-down structure were evaluated using the von-Mises stress and equivalent plastic strain (PEEQ), and one of the ship deck was evaluated using deflection of ship deck and equivalent plastic strain. Compared with each cask drop angle conditions, 45 degree of the cask drop angle showed the highest deflection and PEEQ values, but did not exceed ultimate strain of HT36. In the ship deck, the corner of deck showed the highest PEEQ value in all simulation cases. As the result, the 45 degree of the cask drop angle condition results was more conservative than other conditions, and the corners of deck failure was able to evaluate ship safety.

During normal and off-normal conditions, the concrete structures of dry storage system for spent nuclear fuel must maintain structural integrity. A stress-strain curve is the most important key factor for structural integrity evaluation. The ASTM C39 specifies the concrete specimen geometry for the static compression test. However, there is no standard specimen size for intermediate stain rate, and it is not easy to maintain consistency among all test results because the failure tendency is different from each other. In order to account for the strain rate effects on concrete, the dynamic increase factor (DIF) is conventionally addressed by dividing dynamic strength by static strength. However, the DIF value considers only the strength of concrete and does not describe the overall behavior of concrete, such as a stress-strain relation. The objective of this study is to propose proper specimen geometry for the concrete dynamic compression test by several parametric study. The static compression simulation results with the specimen specified in ASTM C39 showed the constant strain distribution in a cylindrical specimen. However, as the strain rate increases, the strain state in specimen showed a nonuniform with the same geometry of ASTM C39. The non-uniform strain state in the specimen deteriorates the consistency and accuracy of the compression test. Therefore, we presented the specimen shape and size to form a uniform strain state through radial direction by drilling a hole in the axial direction. We analyzed two specimens using ABAQUS with the concrete damaged plasticity model, one with a hole at the center and the other without the hole. As a result, the strain distribution became more uniform than the specimen without the hole. Based on the results, we proposed the specimen shape and size for the intermediate strain rate compression test.

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.

The purpose of this study is to develop the analysis procedures for the evaluation of the structural integrity of the spent fuel in normal condition of transport at sea. Spent nuclear fuel must be transported from the wet storage facility in the nuclear power plant to the intermediate storage facility, and the structural integrity must be maintained in vibration and shock loads during the transportation. In general, the transport of spent nuclear fuel is performed in three kinds of modes: road, rail, and sea. During transport, the spent nuclear fuel is subjected to repeated vibration and shock loads by road surfaces, railroad tracks, and waves of the sea. It should be evaluated whether the structural integrity of the spent fuel is maintained under these load conditions. All nuclear power plants in Korea are located in coastal sites, and the interim storage facility for spent nuclear fuel is highly likely to be decided as a coastal site as well. Therefore, the main mode of the spent nuclear fuel transport is expected to be maritime transport by ships. In this study, the analysis procedure was developed to evaluate the safety of spent fuel at maritime transport by ships, and the procedure for evaluating the integrity of spent fuel under normal conditions of maritime transport were proposed. CFD analysis using SeaFEM was performed for the vibration analysis of the ship by waves, and the structural vibration analysis of the transport system was simulated using the developed in-house codes. The fatigue durability of the cladding was also evaluated using the developed fatigue analysis program and the fatigue analysis used the strain data obtained from the structural analysis. It was concluded that the value of the fatigue damage on the spent fuel cladding during normal conditions of maritime transportation is close to “0” and the structural integrity of the spent fuel is maintained in the same condition.