In this study, we have fabricated the phenolic resin (PR)/polyacrylonitrile (PAN) blend-derived core-sheath nanostructured carbon nanofibers (CNFs) via one-pot solution electrospinning. The obtained core-sheath nanostructured carbon nanofibers were further treated by mixed salt activation process to develop the activated porous CNFs (CNF-A). Compared to pure PAN-based CNFs, the activated PR/PAN blend with PR 20% (CNF28-A)-derived core-sheath nanostructured CNFs showed enhanced specific capacitance of ~ 223 F g− 1 under a three-electrode configuration. Besides, the assembled symmetric CNF28-A//CNF28-A device possessed a specific capacitance of 76.7 F g− 1 at a current density of 1 A g− 1 and exhibited good stability of 111% after 5,000 galvanostatic charge/discharge (GCD) cycles, which verifies the outstanding long-term cycle stability of the device. Moreover, the fabricated supercapacitor device delivered an energy density of 8.63 Wh kg− 1 at a power density of 450 W kg− 1.

In today’s world, carbon-based materials research is much wider wherein, it requires a lot of processing techniques to manufacture or synthesize. Moreover, the processing methods through which the carbon-based materials are derived from synthetic sources are of high cost. Processing of such hierarchical porous carbon materials (PCMs) was slightly complex and only very few methods render carbon nano-materials (CNMs) with high specific surface area. Once it is processed, which paves a path to versatile applications. CNMs derived from biological sources are widespread and their application spectrum is also very wide. This review focuses on biomass-derived CNMs from various plant sources for its versatile applications. The major thrust areas of energy storage include batteries, super-capacitors, and fuel cells which are described in this article. Meanwhile, the challenges faced during the processing of biomass-derived CNMs and their future prospects are also discussed comprehensively.

The amount of waste that contains or is contaminated with radionuclides is increasing gradually due to the use of radioactive material in various fields including the operation and decommissioning of nuclear facilities. Such radioactive waste should be safely managed until its disposal to protect public health and the environment. Predisposal management of radioactive waste covers all the steps in the management of radioactive waste from its generation up to disposal, including processing (pretreatment, treatment, and conditioning), storage, and transport. There could be a lot of strategies for the predisposal management of radioactive waste. In order to comply with safety requirements including Waste Acceptance Criteria (WAC) at the radioactive waste repository however, the optimal scenario must be derived. The type and form of waste, the radiation dose of workers and the public, the technical options, and the costs would be taken into account to determine the optimal one. The time required for each process affects the radiation dose and respective cost as well as those for the following procedures. In particular, the time of storing radioactive waste would have the highest impact because of the longest period which decreases the concentrations of radionuclides but increases the cost. There have been little studies reported on optimization reflecting variations of radiation dose and cost in predisposal management scenarios for radioactive waste. In this study, the optimal storage time of radioactive waste was estimated for several scenarios. In terms of the radiation dose, the cumulative collective dose was used as the parameter for each process. The cost was calculated considering the inflation rate and interest rate. Since the radiation dose and the cost should be interconvertible for optimization, the collective dose was converted into monetary value using the value so-called “alpha value” or “monetary value of Person-Sv”.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

Commercial operation of KORI Unit 1 ended in 2017, and the final decommissioning plan is currently under approval from the KINS. In order for the dismantling waste to go to the repository, it is judged that the radioactive waste generated during the commercial operation should be treated and disposed in advance. Among these radioactive wastes, spent filters contain various radionuclides. The radiation dose rate from the radiation coming out of the filters ranges from a low dose rate to high dose rate. Therefore, in order to handle the spent filters, a remote processing system is required to reduce the radiation exposure of workers. This paper evaluates the radioactive inventory of filters that are stored in the filter room at the KORI unit #1. For this purpose, a method for predicting the radioactivity of each nuclide in the filter, based on the radiation dose rate, has been described using the MicroShield code, which is a commercial shielding code. The information on the filters in the field has only the creation date, type, size, and surface dose rate. In order to evaluate the radioactivity inventory using such limited data, it is possible to know the nuclide radioactivity ratio in the filter. We took out some of the filters stored on site and measured from using the ISCOS system, a gamma nuclide analyzer. The radioactivity of each nuclide in the filter was inferred by modeling with the MicroShield code, based on the radiation dose rate and the radioactivity value of each nuclide measured in the field.

South Korea has been storing UNF in spent fuel pool dry storage facility within Nuclear Power Plants. The dry storage facility of used nuclear fuel (UNF) is essential to sustain safety and sustain stable operation of a nuclear power plant. Most abroad countries have attempted to develop a variety of dry storage facility for used nuclear fuel in order to retain the safe restoration. Many studies have been conducting to safety evaluation for the dry storage facility. However, there is not a ventilation evaluation in the wake of fire event that could influence of the thermal effect on the dry storage facility, even though it will likely to occur fire events such as wildfire, air craft crash. In practice, it happened to catastrophic disaster due to the wild fire adjacent to ul-jin mountain. Also, it happened to fire accident near to the Japonia NPP in Ukraine territory caused of military air plane missile. It has not mostly been studied on the ventilation evaluation considered to thermal safety in the dry storage facility excepted for some researches. It could need the mechanical ventilation systems such as HVAC system in the dry storage system, so that thermal effect can be reduced. In this study, we conducted to the ventilation control modelling by using fire modelling tool (Fire Dynamic Simulator v.6.7). The ventilation scenarios made up for 3 case that can compare flowrate variation with ventilation control. As a result of modelling, there is no differentiation between ventilation control using performance curve with not using performance curve even though the pressure fluctuation would be increased, compared with the case of considering performance curve. Second, it evaluated that the mode for fraction control would occur to pressure rise in the state of controlling the ventilation system flowrate. However, sensitivity of flowrate control was more decreased below less than 5 seconds. Third, in the case of on/off control system revealed more higher resolution than other cases caused by flowrate variation. These results could be considered as the design guidelines for the development dry storage facility to improve the thermal performance that can reduce thermal risk. Furthermore, the study results would expect HVAC system installed in dry storage to help automatic ventilation control relevant to dry storage safety increased.

As regulations on carbon emissions increase, the interest in renewable energy is also increasing. However, the efficiency of renewable energy generation is highly low and has limitations in replacing existing energy consumption. In terms of this view, nuclear power generation is highlighted because it has the advantage of not emitting carbon. And accordingly, the amount of spent nuclear fuel is going to increase naturally in the future. Therefore, it will be important to obtain the reliability of containers for transporting safely and storing spent nuclear fuel. In this study, a method for verifying the integrity and airtightness of a metal cask for the safe transportation and storage of spent nuclear fuel was studied. Non-destructive testing, thermal stability, leakage stability, and neutron shielding were demonstrated, and as a result, suitable quality for loading spent nuclear fuel could be obtained. Furthermore, it is meaningful in that it has secured manufacturing technology that can be directly applied to industrial field by verifying actual products.

Research on the safety of nuclear spent fuel has been heavily experimented and modelled from a mechanical perspective. The issues of corrosion, irradiation creep, hydride and hydrogen embrittlement have been addressed more than two decades since the early 2000s. Among these degradation behavior, hydrogen embrittlement and hydride reorientation have been the most important topics for establishing the integrity of nuclear spent fuel and have been studied in depth. In order to assess the safety of spent nuclear fuel, firstly, it is necessary to establish the safety criteria in all nuclear cycle, i.e., the failure criteria guidelines for nuclear fuel assemblies and nuclear fuel rods, and then examine the safety analysis. The contents of U.S.NRC Regulations, Title 10 General, Chapter 1 Code of Federal Regulation (CFR), Part 50, 71 and 72, describe the safety criteria for the safety assessment of nuclear fuel assemblies and nuclear fuel rods. In this study, technically important points in safety analysis on nuclear fuel are checked through the reference of those NRC regulation. As result, we confirmed that the safety assessment of nuclear fuel after 20 years of interim storage is now being tested by ORNL and PNNL. There are not quantitative criteria related to material safety. However qualitative criteria which is dependent on environmentally condition describe the safety analysis. There is some literature study about DBTT, yield stress, ultimate tensile strength, flexural rigidity data. In FRAPCON code Modelling of yield strength and creep had been established, but radial hydride or hydride reorientation has not considered.

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.

The dry storage of spent fuel has become an increasingly important issue in the field of nuclear energy. Square-gridded baskets have been widely used for the storage of spent fuel because of their superior heat transfer and structural integrity. In this paper, we review the fabrication process of square-gridded baskets for dry storage of spent fuel. The review includes the design considerations, material selection, manufacturing methods, and quality control measures. We also discuss the challenges and opportunities for further improvement in the fabrication of square-gridded baskets. The fabrication of square-gridded baskets is a critical process for the safe and reliable dry storage of spent fuel. The review of the fabrication process highlights the importance of design considerations, material selection, manufacturing methods, and quality control measures. Continued efforts to improve the fabrication process will help to ensure the safe and secure storage of spent fuel.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

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.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

Once systems, structures and components (SSCs) of dry storage systems are classified with respect to safety function or safety significance (i.e., safety classification), appropriate engineering rules can be applied to ensure that they are designed, manufactured, maintained, managed (e.g. aging management) etc. In Unites States, the systems, structures and components (SSCs) consisting DSSs are classified into two or several grades (i.e., class A, B and C or not important to safety, and important to safety (ITS) or not important to safety (NITS)) with respect to intended safety function and safety significance. This classification methods were based on Regulatory Guide 7.10 (i.e., guidance for use in developing quality assurance programs for packaging). Also, in Korea, SSCs of DSSs should be classified into ITS and NITS in much the same as method based on Regulatory Guide 7.10. In that guidance, for providing graded approach to manage the SSCs of packaging, they were trying to classifying SSCs in accordance with radiological consequences. But there was limitations that the provided classification criteria was still qualitative, so that it was not enough for managing the SSCs according to graded approach. On the other hand, in some other nuclear facilities (i.e., nuclear power plant, radioactive waste management facility and disposal facility etc.), quantitative criteria relevant to radiological consequence (i.e., radiation doses to workers or to the public) or inventory of radioactivity are existed so that it can be applied for classifying safety classes. In summary, the study on the application safety classification that applied quantitative criteria to perform safety classification of SSCs in DSS is inadequate or insufficient. The purpose of this study is proposing the preliminary framework for estimating safety significance of SSCs in DSS which can be utilized in our further advanced studies. In this study, a framework was established to estimate the safety significance of SSCs related to radiation shielding and confinement using MCNP® 6.2 and Microsoft Excel. Referring to the methodology of IAEA Specific Safety Guide 30, we assumed severity for failures of components that could lead to degradation of the SSC’s performance. The safety class of SSC was decided based on the impact of SSC’s failure on consequences.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

Al-B4C neutron absorbers are currently widely used to maintain the subcriticality of both wet and dry storage facilities of spent nuclear fuel (SNF), thus long-term and high-temperature material integrity of the absorbers has to be guaranteed for the expected operation periods of those facilities. Surface corrosion solely has been the main issue for the absorber performance and safety; however, the possibility of irradiation-assisted degradation has been recently suggested from an investigation on Al-B4C surveillance coupons used in a Korean spent nuclear fuel pool (SFP). Larger radiation damage than expectation was speculated to be induced from 10B(n, α)7Li reactions, which emit about a MeV α-particles and Li ions. In this study, we experimentally emulated the radiation damage accumulated in an Al-B4C neutron absorber utilizing heavy-ion accelerator. The absorber specimens were irradiated with He ions at various estimated system temperatures for a model SNF storage facility (room temperature, 150, 270, and 400°C). Through the in-situ heated ion irradiation, three exponentially increasing level of radiation damages (0.01, 0.1, and 1 dpa or displacement per atom) were achieved to compare differential gas bubble formation at near surface of the absorber, which could cause premature absorber corrosion and subsequential 10B loss in an SNF storage system. An extremely high radiation damage (10 dpa), which is unlikely achievable during a dry storage period, was also emulated through high temperature irradiation (350°C) to further test the radiation resistance of the absorber, conservatively. The irradiated specimens were characterized using HR-TEM and the average size and number density of radiation-induced He bubbles were measured from the obtained bright field (BF) TEM micrographs. Measured helium bubble sizes tend to increase with increasing system (or irradiation) temperature while decrease in their number density. Helium bubbles were found from even the lowest radiation damage specimens (0.01 dpa). Bubble coalescence was significant at grain boundaries and the irradiated specimen morphology was particularly similar with the bubble morphology observed at the interface between aluminum alloy matrix and B4C particle of the surveillance coupons. These characterized irradiated specimens will be used for the corrosion test with high-temperature humid gas to further study the irradiation-assisted degradation mechanism of the absorber in dry SNF storage system.

Currently, in the United States, Spent Nuclear Fuel (SNF) is stored at the Independent Spent Fuel Storage Installations (ISFSIs) at 73 Nuclear Power Plants (NPPs). The SNF inventory stored on-site either in pools or dry storage was 84,500 MTU in 2020. The inventory stored in on-site dry storage facilities was 39,207 MTU (46% of the total), and it is growing at a rate of approximately 3,500 MTUs per year. However, because a site for geologic repository for permanent disposal of SNF has not been constructed in the U.S., the SNF will need to be stored in dry storage facilities across the U.S. for a much longer period of time than originally planned. During this time, the dry storage facilities could experience earthquakes of a different magnitude than the one for which they were originally designed. However, there is little data on the response of SNF inside dry storage systems to seismic loads in the U.S., and the various gaps and nonlinearities between storage containers, canisters, baskets, aggregates, and fuel make it very difficult to evaluate by analytical methods. Therefore, a full-scale shake table test is being planned as an international joint research project led by Sandia National Laboratories (SNL) in the U.S. In Korea, KNF decided to participate in this seismic test through the project of SNF integrity evaluation under road and sea normal transportation conditions organized by KNF and conducted by KORAD, KAERI, and Kyung-Hee University, and has provided the KNF 17ACE7 and PLUS7 test assemblies for the tests to SNL. The test will be conducted at the LHPOST6 shake table test facility operated by University of California in San Diego (UCSD) from 2023 to 2024, with the participation of KNF, CRI, and KAERI in Korea. The test units consist of a NUHOMS 32 PTH2 canister, a mockup of a generic vertical cask, a mockup of a generic horizontal storage module, 4 surrogate fuel assemblies, and 28 dummy assemblies. The seismic inputs for the tests will consist of ground motions (acceleration time histories) representative of hard rock, soft rock, and soil sites and seismic conditions in moderately tectonically active Central and Eastern US and highly tectonically active Western US. Ground accelerations for soft rock and soil conditions will be developed taking in account soil-structure interaction. Not only is this test almost impossible to conduct independently in Korea in terms of scale, facilities and costs, but it is also considered an essential test for those of us who are preparing for dry storage of spent nuclear fuel, given the increasing social concern about earthquakes due to the recent earthquake in Turkey.