While many countries consider direct disposal of the spent nuclear fuels, they need to consider long-term disposal scenarios with severe accidents such as the contact between underwater and the spent nuclear fuel due to large defect of the canister. Radionuclides releases rapidly with contacting water or slowly with dissolution of UO2 matrix. The former is known as the ‘Instant Release’, and the latter is ‘Congruential Release’. Even though the instant release fractions (IRF) are much smaller than the congruential ones, IRF has to be treated carefully due to the fact that the instant releases lead to much larger value of the exposure dose rates than the congruential ones which proceed very slowly. It is known that the exposure dose rates by the instant releases are ~25 times larger than the one by the congruent release. The radionuclides from UO2 matrix migrate to the grain boundary, make bubbles, and make tunnels, which leads to instant releases of some radionuclides. The radionuclides in the gap between UO2 pellet and cladding can be also instantly released. In addition, the radionuclides in the crud are instantly released. But in this paper, nuclides from the crud are not regarded, due to the lack of the leaching data. Meanwhile, there’re some nuclides that released from the construction materials like the cladding, the Rod Cluster Control Assembly (RCCA), or the other metal parts. In this work, IRF values for major IRF nuclides such as Cs, I, Cl, Se for the reference PWR spent fuels of South Korea were evaluated based on the rationale from literatures’ review. In particular, these evaluations were done as the function of fission gas release (FGR), average discharge burnup, and fuel dimensions. In addition, the values of IRF for the other nuclides were also suggested based on the other institutes.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.

This study examined the heat balance in the electrolytic reducer during oxide reduction of pyroprocessing. The adoption of carbon anodes instead of conventional platinum anodes in the oxide reduction process has made it possible to apply high currents, and it has been observed that the temperature of the molten salt of in the reactor rises rapidly when applying high currents, so it is important to maintain an optimal operational temperature range. In this study, salt resistant heat, reaction heat, and decay heat were identified as factors affecting heat balance during the operation of oxide reduction process. Equations describing the relationships among these factors were established. Then using this, a correlation was developed to understand the relationship between applied current and the molten salt temperature in the reactor observed in the actual operation of the carbon anode electrolytic reducer of KAERI. Furthermore, this study proposed strategies to mitigate excessive temperature elevation during oxide reduction operation. A comparative assessment of these approaches was conducted. Considering KAERI electrolytic reducer operation environment, among the considered cooling strategies, the cooling effectiveness was calculated to be highest in the following order: heat transfer to extra salt, convection, conduction, argon gas bubbling.

Safe management of spent nuclear fuel (SNF) is a key issue to determine sustainability of current light water reactor (LWR) fleet. However, none of the countries are actually conducting permanent disposal of SNFs yet. Instead, most countries are pursuing interim storage of spent nuclear fuels in dry cask storage system (DCSS). These dry casks are usually made of stainlesssteels for resistibility against cracking and corrosion, which can be occurred over a long-term storage period. Nevertheless, some corrosion called Chloride-Induced Stress Corrosion Cracking (CISCC) can arise in certain conditions, exacerbating the lifetime of dry casks. CISCC can occur if the three conditions are satisfied simultaneously: (i) residual tensile stress, (ii) material sensitization, and (iii) chloride-rich environment. A residual tensile stress is developed by the two processes. One is the bending process of stainless-steel plates into a cylindrical shape, and the other is the welding process, which can incur solidification-induced stress. These stresses provide a driving force of pit-to-crack transition. Around the fusion weld areas, chromium is precipitated at the grain boundary as a carbide form while it depletes chromium around it, leading to material susceptible to pitting corrosion. It is called sensitization. Finally, coastal regions, where nuclear power plants usually operate, tend to have a higher relative humidity and more chloride concentration compared to inland areas. This high humidity and chloride ion concentration initiate pitting corrosion on the surface of stainless-steels. To prevent initiation of CISCC, at least one of the three conditions should be removed. For this, several surface engineering techniques are under investigation. One of the most promising approaches is surface peening method, which is the process that impacts the surface of materials with media (e.g., small pins, balls, laser pulse). By this impact, plastic deformation on the surface occurs with compressive stress that counteracts with pre-existing residual tensile stress, so this approach can prevent pit-to-crack transition of stainless-steels. Also, cold spray deposition can prevent CISCC. Cold spray deposition is a method of spraying fine metal powder to a substrate by accelerating them to supersonic velocity with propellant gas. As a result, a thin coating composed of the feedstock powders can protect the substrate from outer corrosive environments. In addition, the impact of the feedstock powder on the substrate during the process provides compressive stress, similar to the peening method.

Globally, the operation of nuclear power plants results in the production of a tremendous quantity of spent nuclear fuel. The methods for handling spent nuclear fuel can be categorized into three: storage, direct disposal and recycling. A technology designed to recycle accumulated spent nuclear fuel is pyropocessing. In pyroprocessing, various fission products (FPs) such as C-14, H-3, I-129 and Cs-137 are generated. Among these FPs, technetium (Tc-99) is a gaseous nuclear isotope with a long half-life and high mobility in the form of TcO4 - in aqueous solutions, making it essential to capture strictly in order to prevent radioactive contamination of the environment. In previous studies, ion-exchange or adsorption using MOFs (Metal Organic Frameworks) have been used to remove Tc-99. These methods, however, involve separation in aqueous solutions, not in the gaseous state. In this study, we developed a CaO-based adsorbent for capturing Re as a surrogate for radioactive Tc-99. Isopropyl alcohol (IPA) was employed as a pore-forming agent during the preparation of the adsorbents, and its effects on characteristics and adsorption performance were investigated. The size of the pores were analyzed from nitrogen (N2) adsorption isotherm analysis and mercury (Hg) intrusion curves. As a result, it was confirmed that the addition of IPA had a significant impact on the formation of macro-pores. Furthermore, this macroporous structure was found to enhance the adsorption performance of Re.

It has been investigated on the management of Strontium-90 in KAERI. It is needed to separate the solute from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt in our previous study. Strontium chloride was successfully carbonated by reactive distillation of SrCl2 – K2CO3 – LiCl – KCl system. In this study, it was tried to develop another route to recover strontium from the salt solution by a solid-solid reaction for avoiding the entrainment of product and the salt-K2CO3 reaction. Reactive distillation experiments were carried out for SrCl2 - K2CO3 – LiCl – KCl system. The carbonation temperature and pressure were 520°C and 0.8 bar. After the carbonation reaction, the temperature was elevated to 820°C to remove KCl from the reaction product. SrCO3 and KCl peaks were found in the XRD analysis of the residual product. It could be concluded that SrCl2 can be successfully carbonated after salt removal by the solid-solid reaction.

NFDC (Nuclear Fuel and materials Data Center) is designated as a one of the data center of National Standard Reference Center from Ministry of Trade Industry and Energy at Dec. 30 2008. The fields of designation were nuclear fuel and energy materials. NFDC produces standard reference data of nuclear fuel and materials. To ensure reliability of experimental data uncertainty should be estimated. There are two kinds of uncertainty: A-type uncertainty from tester and B-type uncertainty from experimental equipments. To reduce the former, the measurement should be repeated for sufficient amount of times, and to reduce the latter type uncertainty all equipment have to be calibrated. In this study self calibration process of thermo-mechanical analyzer (TMA) was established to ensure the B-type uncertainty. The self calibration was performed using the standard reference material and correction factor was obtained. The correction factor was defined as the ratio of the thermal expansion value of the standard reference material reported in the certificate and the thermal expansion value measured using TMA. It is believed that the uncertainty evaluation process of TGA data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

Various disposal methods for spent nuclear fuels (SNFs) are being researched, and one of these methods involves separating high heat-generating nuclear isotopes such as Strontium-90 (90Sr) and Cesium-137 (137Cs) for deep disposal. These isotopes has relatively short half-lives and substantial decay energies. Especially, 90Sr undergoes decay through Yttrium-90 to Zirconium-90, emitting intense heat with beta radiation. Therefore, the removal of these high heat-generating isotopes will significantly contribute to reducing disposal site area. To remove 90Sr from SNFs, molten salt was utilized in KAERI. During this process, it was discovered that 90Sr dissolves in the molten salt in the form of SrCl2 and/or Sr4OCl6. Afterwards, it is crucial to recover 90Sr in the form of oxide from the salt to create immobilized forms for disposal. This can be achieved by reactive distillation with K2CO3. However, the amount of 90Sr within the SNFs is only 0.121wt%, and even if all the 90Sr in the SNFs were to leach into the molten salt, the quantity of 90Sr in the molten slat would still be very small. Therefore, adding K2CO3 to the molten salt for reactive distillation could result in significant possibilities of side reactions occurring. In this study, a two-step process was employed to mitigate the side reactions: the 1st step involves evaporating the all molten salts and the 2nd step includes adding K2CO3 to make oxides through solid-solid reaction. Eutectic LiCl-KCl, which is the most commonly used salt, was employed. The eutectic LiCl-KCl with SrCl2 was heated at 850°C for 2 h to evaporate the salts under a vacuum (> 0.02 torr). However, after examining the distillation product before the solid-solid reaction, it was observed that SrCl2 reacted with KCl in the salt, resulting in the formation of KSr2Cl5. It means that salts containing KCl are not suitable candidates for reactive distillation aimed at producing immobilized forms. As an alternative, MgCl2 could be a highly promising candidate because it is inert to SrCl2 and according to a recent study in KAERI, MgCl2 exhibited the most efficient separation of Sr among various salts. Therefore, we plan to proceed with the two-step reactive distillation using MgCl2 for the future work.

In KNF, fuel performance analysis modules were developed to predict the overall behavior of a fuel rod under normal operating conditions. Their main focus is to provide information on initial conditions prior to dry storage. Potential degradation mechanisms that may affect sheath integrity of spent CANDU fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed modules that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules, identify and extend the ranges of all modules to required operating ranges. The 300°C spent CANDU fuel sheath temperature metric for dry storage ensures spent CANDU fuel element integrity from the failure mechanisms of creep rupture, oxidation and stress corrosion cracking at a failure probability of 2×10-5 for a dry storage time of 100 years. The 300°C sheath temperature metric for dry storage has relatively a lower failure rate than the target criteria for dry storage of spent LWR fuel. Although different modes of failure were treated separately for simplicity, ignoring possible synergistic effects, these results are conservative because of the conservative assumptions that have been made for evaluating spent fuel element conditions, and because of the inherent conservatism of the applied models. Additional conservatism of the model comes from the fact that isothermal conditions do not prevail in actual storage conditions. Further R&D being considered includes acquisition of new functional models to implement overall fuel behavior evaluation and cover spent CANDU fuel in dry storage, and upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The developed modules provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for spent CANDU fuel.

Molten salt reactor (MSR) uses fluoride or chloride based molten salt as a coolant of the system, and fuel materials are dissolved in the molten salt, therefore it can be act as both coolant and nuclear fuel. A few issues have arisen from early-stage research and development program of MSR from Oak Ridge National Laboratory, including corrosion of structural materials and fission product management. For investigating the effect of additives on corrosion of structural materials, Mg(OH)2 and MgCl2*6H2O are added into the NaCl-MgCl2 eutectic salt. Prepared chloride salt is injected into the autoclave in the glove box, as well as corrosion coupons for candidate structural materials for molten chloride salt reactor, SS316, Alloy 600, and C-276 are also prepared. The temperature is set as 700°C. After 500 h corrosion experiment, the samples are taken out from the autoclave, and they are analyzed with scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS). SS316 samples show weight loss with all salt conditions, while Alloy 600 and C-276 show weight gain after the corrosion experiment.

An austenitic stainless steel canister functions as a containment barrier for spent nuclear fuel and radioactive materials. The canister on the spent fuel storage system near the coastal area has several welding lines in the wall and lid, which have high residual tensile stresses after welding procedure. Interaction between austenitic stainless steel and chloride environment from a sea forms a detrimental condition causing chloride induced stress corrosion cracking (CISCC) in the canister. The South Korea is concerned with the dry storage of high-level spent nuclear fuel and radioactive wastes to be built on the site of a nuclear power plant. The importance of aging management has recently emerged for mitigating CISCC of dry storage canisters. When a corrosive pit is created by a localized corrosion in a sea water atmosphere, it initiates and grows as CISCC crack. Surface stress improvement works by inducing plastic strain which results in elastic relaxation that generates residual compressive stress. Surface stress improvement methods such as roller burnishing process can effectively mitigate the potential for CISCC of the canister external surfaces. The generation of compressive stress layer can inhibit the transition to cracking initiation. In this study, a flat roller burnishing process was applied as a prevention technology to CISCC of stainless steel canisters. Roller burnishing process parameters have been selected for 1:3 scale canister model having a diameter of 600 mm, a length of 1,000 mm and a thickness of 10 mm on the basis of the burnishing conditions available to control residual tensile stress of austenitic stainless steel plate specimens. The surface roughness of the scaled canister model was investigated using a surface roughness measurement equipment after roller burnishing treatment. The surface residual stresses of the scaled canister model were measured by a hole drilling contour method attached with strain gauge. The burnishing test results showed that the surface roughness of the scaled canister model was considerably improved with flat rollers having the tip width of 4 mm. The surface of the scaled canister model had significant residual compressive stress after burnishing treatment. The roller burnished canister with good surface roughness could reduce the number of crack initiation sites and the residual compressive stress formed on the welded surface might prevent the crack initiation by reducing tensile residual stress in the weld zone, finally leads to CISCC resistance.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

It has been known that as oxide layer (ZrO2) forms on the nuclear fuel cladding during irradiation in nuclear power plants, the corrosion kinetics are influenced by various parameters such as chemical environments. One of those environments, crud deposition driven by coolant chemistry has an adverse effect on the formation of oxide (ZrO2) and leads to increase thickness of the layer. In this study, crud formation was performed through loop experiment equipment on the surface of intentionally-made oxide layer (ZrO2) on cladding tubes and then the composition and characteristics of cruds were examined for the investigation of nuclear power plant environment. As a result, various cruds in composition and microstructure were formed depending on the exquisite methods and conditions such as metal ion concentration.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

Pyroprocessing is a crucial method for recovering nuclear fuel materials, particularly uranium and transuranic elements (TRU), through electrochemical reactions in a LiCl or LiCl-KCl molten salt system, which is highly stable medium at elevated temperatures. In the electrochemical reduction stage, actinide metal oxides are effectively transformed into their metallic forms and retained at the cathode within a molten LiCl-Li2O environment at 650°C. Simultaneously, oxygen ions (O2-) are generated at the cathode and then transported through the molten salt to be discharged at the anode, where they combine to form oxygen gas (O2) on the anode’s surface. One notable challenge in this electrochemical process is the generation of various byproducts during the anode oxide reduction step, including oxygen, chlorine, carbon dioxide, and carbon monoxide. Consequently, significant amounts of corrosion products tend to accumulate on the upper region of the anode’s immersion area over time. This report introduces a novel solution to mitigate corrosion-related challenges within the specified temperature range. We propose a selective oxidation treatment for the NiCrAl-based 214 Haynes alloy, involving exposure to 1,100°C in a reducing atmosphere. The objective is to stimulate the growth of protective α-Al2O3 scales on the alloy’s surface. The resulting oxide scales have undergone thorough characterization using SEM, EDS, and XRD techniques. The pre-grown alumina scale has demonstrated commendable adherence and thermal stability, even when subjected to a chlorine-oxygen mixed atmosphere at the specified temperature.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

More than 20,000 bundles of spent nuclear fuel are stored in the spent nuclear fuel storage pool of domestic nuclear power plants, and the dry storage facility project in the nuclear power plant site is being promoted as the saturation of the wet storage pool is imminent. Since bending or twisting of spent nuclear fuel is an important item in order to load spent nuclear fuel into a dry storage cask, PSE (Pool Side Examination) was performed to verify this. This paper describes whether it can be safely loaded into a dry storage cask based on the measurement results of bending or twisting of spent nuclear fuel. The nuclear fuel assembly is designed to prevent excessive assembly bending and twisting because it can cause interference during dry storage and handling due to factors such as differences in depletion of nuclear fuel rods, irradiation growth, and coolant flow during reactor operation. The bending of the nuclear fuel assembly is measured by establishing a Plumb Line to photograph the nuclear fuel assembly based on it, and calculating a pixel that images the distance between the support grid and the Plumb Line. The twisting of the nuclear fuel assembly is measured by forming a virtual vertical plane with two Plumb Lines, and based on this, the twisting angle of the lower fixed compared to the upper fixed. As a result of the measurement, the bending of spent nuclear fuel was about 0.0-10.2 mm, much lower than the reactor loading criteria of 15.0 mm, and in the case of twisting, about 0.0~2.2° much lower than the reactor loading criteria of 5.0°. Therefore, it was confirmed that spent nuclear fuel at domestic nuclear power plants was not affected by bending and twisting when loading into dry storage cask.

The separation efficiency of nuclides in molten salt systems was investigated, with a focus on the influence of apparatus configuration and experimental conditions. A prior study revealed that achieving effective Sr separation from simulated oxide fuel required up to 96 hours, reaching a separation efficiency of approximately 90% using a static dissolution reaction in a porous alumina basket. In this study, we explored the impact of agitation on improving Sr separation efficiency and dissolution rates. The simulated oxide fuel composition consisted of 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. To quantify the Sr concentration in the salt, we utilized ICP analysis after salt sampling via a dip-stick technique. Furthermore, we conducted ICPOES analysis over a 55-hour duration to assess the separated nuclides. Complementing these analyses, SEM and XRD investigations were performed to validate the crystal structure and morphology of the oxide products.

The potential use of cost-effective carbon anodes, as an alternative to expensive platinum, in the reduction of oxides within LiCl-Li2O molten salt at elevated cell potentials presents a promising avenue. However, this elevated potential gives rise to the generation of a complex mixture of anodic gases, including hazardous and corrosive species such as chlorine (Cl2), oxygen (O2), carbon monoxide (CO), and carbon dioxide (CO2). In this study, we investigate the influence of applied potential and salt composition on the composition of the generated gas mixture. Real-time gas analysis was conducted during the TiO reduction reaction in the molten salt at 650°C using a MAX-300-LG gas analyzer. Simultaneously, electronic signals, including current, potential, and salt composition, were monitored throughout the oxide reduction process. Additionally, XRD investigations were performed to verify the crystal structure of the resulting products. This research provides valuable insights into optimizing carbon anode-based reduction processes for improved efficiency and safety.