The spent fuel is classified based on the arrangement of fuel rods, which is considered the primary characteristic data for selecting nuclear fuel. The reason for prioritizing the classification by fuel rod arrangement is that it has the greatest physical impact on the production, supply, operation, reactor type, rack size within the containment vessel, and specifications for the basket in the future dry storage system. Additionally, as mentioned earlier, various meanings of nuclear fuel types are distinguished according to the arrangement of fuel rod. The burnup and cooling period ranges are also important factors in the characterization analysis for the selection of spent fuel, the burnup range was set for both low and high burnup ranges and the cooling period is necessary to consider the reliability during handling of nuclear fuel thermal distribution within the storage system

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.

As if the wet storage of Spent Nuclear Fuel (SNF) becomes saturated, a transition from wet storage to dry storage could be required. The first process for dry storage is to move SNF from the wet storage into a canister for dry storage, and secondly perform a drying process to remove the moisture in the canister to prevent a potential impact such as deterioration of cladding or corrosion of the interior material. Nuclear Regulatory Commission (NRC) accepts the conditions describing the adequate dryness state that remain below the pressure of 3 Torr for 30 minutes in the drying process. That is, the most pressure of water vapor that may exist inside the canister is 3 Torr. If it is maintained below 3 Torr, it can be determined that the dryness criterion is satisfied. Based on this, relative humidity and dew point trends can be identified. Relative Humidity (RH) is calculated by dividing the vapor pressure by the saturated vapor pressure. Here, if the vapor pressure is fixed at 3 Torr, which is the dryness criterion value, the relative humidity has a value according to the saturated vapor pressure. Saturated vapor pressure is a value that varies with temperature, so relative humidity varies with temperature. On the other hand, the dew point temperature has a value according to the water vapor pressure. Therefore, when the internal temperature of the canister is 120°C and the water vapor pressure is 3 Torr, the relative humidity is 0.2% and the dew point temperature is -4.4°C. We will confirm the suitability of the dryness criterion through the drying tests, and secure a technology that can measure and evaluate the amount of moisture remaining inside the canister.

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.

As of 2023, there has been significant progress worldwide in the management of nuclear fuel’s spent radioactive waste (HLW). Several countries have made important strides in advancing their plans for the construction of deep geologic repositories (DGRs) to safely dispose of their nuclear waste. Finland led the way, with its nuclear waste management organization, Posiva Oy, submitting an application for an operating license for a DGR for spent fuel generated by the nuclear power plants of its owners. The facility, ONKALO, will be located on the island of Olkiluoto and is expected to begin final disposal in the mid-2020s. Sweden also approved SKB’s application to build a DGR in Forsmark, and an encapsulation plant next to the Clab interim storage facility. In Switzerland, Nagra selected Nordic Lagern as the site for the Swiss DGR, and is preparing the general license applications for the required facilities. Meanwhile, Canada’s Nuclear Waste Management Organization (NWMO) narrowed down the possible locations for its DGR to two, and expects to name its preferred site by fall 2024. The UK established four Community Partnerships to participate in the siting process for a DGR, with Nuclear Waste Services (NWS) responsible for identifying a site. Andra, the French organization responsible for managing all French radioactive waste, is expected to submit an application by the end of the year for a DGR in France that will contain HLW resulting from reprocessing of spent fuel assemblies from French nuclear power plants, as well as intermediate-level waste. Overall, the progress made by these countries represents a tangible and sustainable step forward in the management of spent fuel and HLW, and brings us closer to the safe and effective long-term disposal of nuclear waste.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).

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.

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.

The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating the thermal safety of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The management of spent nuclear fuel function provides a management tool for spent nuclear fuel in the SFP. The fuel assemblies both in SFP and reactor side can be shown graphically in the screen. The loading sequence into transfer cask can be checked respectively in the CAREPOOL. A basic heat balance equation was used to estimate the SFP temperature using the heat load calculated in the previous step. The characteristics of typical SFPs and associated cooling systems at reactor sites in the Korea were applied. Accident simulation like station black out leading to loss of SFP cooling or inventory is possible. Emergency cooling water injection pipe installed subsequent to the events at Fukushima 2011 is also modeled in this system. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, fuel selection for periodic spent fuel transferring campaign. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation.

Flow-induced vibration can lead to fretting wear damage of fuel rods and spacer grids in nuclear reactors. Similarly, during the transport of spent nuclear fuel assemblies, continuous vibration and intermittent impact might also result in fretting wear due to dynamic interaction. Therefore, it is important to evaluate fuel rod damage due to fretting wear under such transport conditions. This study examines spent nuclear fuel rod specimens fabricated with hydride cladding tubes and simulated pellets, with hydrogen content ranging from 200 to 700 ppm and oxide film thickness ranging from 25 to 100 micrometers. Tests were conducted under a worst-case scenario, assuming continuous exposure to that condition during the expected transport time. Results indicate that the wear depth of all rod specimens occurred within the oxide film, suggesting a high resistance to fretting wear during transportation.

CANDU Spent Fuel (CSF) dry storage system, SILO, has been operated from 1992 at Wolsung under 50 year operating license. As of 2023, this system has been operated for over 30 years and its licensed remaining operation time is less than 20 years. When it faces the final stage of operation, it has only two options; moving to a centralized away-from-reactor storage or extending its license atreactor. These two options have an inevitable common duty of confirming the CSF integrity by a “demonstration test”. Since the degradation of CSF and structural materials in the SILO are critically dependent on temperature, two important goals of the ‘DEMO test’ were set as follows. 1. Design of ‘DEMO SILO’: Development of internal monitoring technology by transforming SILO design. 2. Accurate measurement and evaluation of the three-dimensional temperature distribution in the ‘DEMO SILO’ Based on operating real commercial SILO dimension, a conceptual “DEMO SILO” design has been developed from 2022. Because, unlike with commercial Silo, ‘Demo Silo’ must be disassembled and assembled, and have penetration holes. Safety evaluation technologies like structural, thermal and radiation protection analysis also have been developed with design work. ‘Demo SILO’ should evaluate an accurate 3D temperature distribution with minimal number of thermocouples and penetration holes to avoid disruption of internal flow and temperature distribution. For this reason, a ‘Best Estimate Thermal-Hydraulics evaluation system for SILO’ is under development and it will be essential for ensuring temperature prediction accuracy. Construction of a full-scale test apparatus to validate this technology will begin in 2024. In order to supply power to many heaters and monitor temperature gradient inside of this apparatus, it has modular design concept by dividing its whole body to axial 9 sub-bodies which looks like a donut containing a basket at center position.

Spent nuclear fuel (SNF) characterization is important in terms of nuclear safety and safeguards. Regardless of whether SNF is waste or energy resource, the International Atomic Energy Agency (IAEA) Specific Safety Guide-15 states that the storage requirements of SNF comply with IAEA General Safety Requirement Part 5 (GSR Part 5) for predisposal management of radioactive waste. GSR Part 5 requires a classifying and characterizing of radioactive waste at various steps of predisposal management. Accordingly, SNF fuel should be stored/handled as accurately characterized in the storage stage before permanent disposal. Appropriate characterization methods must exist to meet the above requirements. The characterization of SNF is basically performed through destructive analysis/non-destructive analysis in addition to the calculation based on the reactor operation history. Burnup, Initial enrichment, and Cooling time (BIC) are the primary identification targets for SNF fuel characterization, and the analysis mainly uses the correlation identified between the BIC set and the other SNF characteristics (e.g., Burnup - neutron emission rate) for characterizing. So further identification of the correlation among SNF characteristics will be the basis for proposing a new analysis method. Therefore, we aimed to simulate a SNF assembly with varying burnup, initial enrichment, and cooling time, then correlate other SNF properties with BIC sets, and identify correlations available for SNF characterization. In this study, the ‘CE 16×16’ type assembly was simulated using the SCALEORIGAMI code by changing the BIC set, and decay heat, radiation emission characteristics, and nuclide inventory of the assembly were calculated. After that, it was analyzed how these characteristics change according to the change in the BIC set. This study is expected to be the basic data for proposing new method for characterizing the SNF assembly of PWR.

The current storage capacity of the spent nuclear fuel at the Kori unit 2 has reached approximately 94% saturation, excluding emergency core capacity. As South Korea has not yet constructed any interim storage facilities to store spent nuclear fuel, one of possible options is to install high density storage racks in spent fuel pool at the reactor site to increase its capacity. The high density storage rack is a technology developed to allow the storage to have more spent nuclear fuel than originally planned, while still ensuring safety. It achieves this by reducing the storage gap between the spent nuclear fuel. The purpose of this study is to investigate an appropriate storage capacity for spent fuel pool in the Kori unit 2 and the factors to be considered during the high density storage rack installation. Given that the Kori unit 2 is planning continued operation and Korea Hydro & Nuclear Power (KHNP) is preparing to install a temporary storage facility for spent nuclear fuel at the Kori nuclear site, it is reasonable to consider the installation of high density storage racks in the spent fuel pool as a storage solution for these issues. When evaluating the capacity of the spent fuel pool, the amount of spent nuclear fuel generated by other reactors in Kori nuclear site and the amount of spent nuclear fuel generated by continued operation of the Kori unit 2 should be taken into account. This study aims to consider these factors and evaluate the capacity of the spent fuel pool. Furthermore, when installing high density storage rack for the spent nuclear fuel, it should be noted that the existing storage racks at the Kori unit 2 are welded to the liner plate, which may require additional cutting work. Therefore, it is necessary to review the suitable method for the cutting work. Additionally, assuming that divers need to access the area near the storage racks or cutting & welding devices require radiation protection in the area, it is essential to analyze the expected radiation level with computational code and propose appropriate measures to limit work time or establish a work zone. Thus, this study evaluates appropriate capacity of spent fuel pool and work methods for the installation of high density storage rack in the spent fuel pool at the Kori unit 2. It is expected that this paper contributes to install high density storage racks in SFP safely.

To improve the safety of nuclear fuel, research on the advanced nuclear fuel (UO2) by adding various trace elements is being conducted. For example, the addition of metals such as Mo, Cr can improve the thermal conductivity of nuclear fuel, minimizing the diffusion of fission products. Trace metal oxide additives (SiO2, Cr2O3, Al2O3, etc.) can suppress the release of fission gases. In general, complete dissolution of the fuel sample is required for chemical analysis to determine its elemental compositions. Among widely used metal oxide additives, aluminum oxide is difficult to dissolve in nitric acid due to its excellent thermal and chemical stability. In this study, we investigated on different chemical dissolution methods by applying a microwave digestion system under various acid solutions. We confirmed the validity of the digestion method by carrying out trace element analysis using an Inductively-Coupled Plasma Atomic Emission Spectrometer (ICP-AES).

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

In this paper, a basic study was conducted to observe the temperature inside the tube according to the heating temperature of the tube furnace. In a tube furnace, a tube is inserted, and the air space outside the tube is heated to increase the temperature of the gas inside the tube through conduction of the tube. Tube furnaces are widely used in research to capture volatile nuclides. In this case, a volatile nuclide capturing filter is inserted inside the tube, and an appropriate temperature is required to capture it. Since the tube furnace heats the air space outside the tube to the target temperature, a difference from the temperature inside the tube occurs. In particular, if a flow of gas occurs inside the tube, a larger temperature difference may occur. In order to confirm this temperature difference, an experimental device was constructed, and basic data was produced through several experiments. The following studies were conducted to produce data. First, the temperature of the air layer of the heating unit and the temperature inside the tube were measured in real time in the absence of gas flow inside the tube. Second, the temperature of the air layer of the heating unit and the temperature inside the tube were measured in real time while air having a certain temperature was flowing inside the tube. As a result of the experiment, when there is no flow inside the tube, when the heating target temperature is low, the temperature inside the tube is significantly lower than the target temperature, and when the target temperature is high, the temperature inside the tube approaches the target temperature. It was found that when there is about 20°C air flow inside the tube, the temperature inside the tube is significantly lowered even if the heating target temperature is high. In the future, additional research on changing the temperature of the gas flowing inside the tube will be conducted, and the results of this study are expected to greatly contribute to the design of a tube furnace that captures volatile nuclides.

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 stabilization technology for the damaged spent fuel is being developed to process the damaged fuel into sound pellet suitable for dry re-fabrication. It requires several treatments including oxidative decladding followed by reduction treatment for oxidized powder closely related to the quality of oxidized powders for pellet fabrication. For the development of operating condition for the reduction treatment, in this study, we evaluated the effect of air-cylinder based vertical shaking previously applied to oxidative decladding on powder reduction. For U3O8 of 50-100 g, the reduction test were applied with and without vertical shaking at 700°C under reduction atmosphere (Ar + 4%H2) and the concentration of hydrogen in effluent was measured to evaluate the reduction reaction. It was found that the vertical shaking system has allowed the reaction time of 50 g and 100 g U3O8 reduced by 33% compared to the test in static mode. Based on XRD analysis, the better crystallinity of the products was also achieved.

Spent Fuel Pool Island (SFPI) is a spent nuclear fuel storage pool that operates independently of existing nuclear facilities to safely manage SNF and minimize maintenance costs during the nuclear decommissioning process. Since the radiation controlled area can be dismantled before transporting SNF to a dry storage facility, the overall decommissioning period can be shortened, and the risk of occupational exposure during dismantling is reduced. In the US, various nuclear power plants have introduced SFPI for this reason. In this paper, to analyze the economic feasibility of application of SFPI to nuclear power plants to be decommissioned, several scenarios are established in consideration of the decommissioning plan and schedule, SFPI and dry storage facility application schedule. Cost and benefit list (SFPI application cost, SNF management cost, SNF dry storage cask cost, etc.) for each alternative were derived, and economic analysis was conducted by applying the Net Present Value (NPV). As a result of the analysis, it is found that the application of SFPI during decommissioning is economically effective as the NPV showed a positive number even when uncertainty was taken into account.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. Finally, the shake table tests and rolling test were conducted from October 31 to November 2, 2022. The shake table test was performed with the input load produced conservatively from the data obtained from the road and sea transportation tests. The test input was produced based on the power spectral densities and shock response spectrums from the transportation tests. In addition to the test inputs from the road and sea tests, sine sweep input and half sine input were used to verify the vibration characteristics of assemblies under boundary conditions during normal conditions of transport. Because the input load of the shake table test was produced conservatively, a slightly larger strain than the strain value measured in road and sea transportation tests was measured from the shake table tests. In the case of the sea test, it is considered that the process of enveloping the data in the 20 to 80 Hz range generated by the engine propeller system was performed excessively conservatively. As a result of analyzing the test results for the difference in boundary conditions, it was confirmed that the test conditions of loading the basket generated a relatively large strain compared to the conditions of loading the disk assembly for the same input load. Therefore, it is concluded that a transportation cask having a structure in which a basket and a disk are separated, such as KORAD-21, is more advantageous in terms of vibration shock load characteristics under normal conditions of transport than a transportation cask having an integral internal structure in which a basket and a disk are a single unit. However, this effect will be insignificant because the load itself transmitted to the disk assembly is very small.