Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.

The concept of deep geological disposal for high-level radioactive waste is based on an engineered barrier system (EBS), including a canister, bentonite buffer and backfill material. The bentonite buffer is key component of the EBS to prevent groundwater infiltration and radionuclide leakage. However, the bentonite buffer can become saturated due to groundwater flow through the excavation damaged zone in the adjacent rock, causing erosion of bentonite buffer and affecting the long-term performance of EBS. While the RH (relative humidity) sensor is commonly used to assess the degree of saturation in the bentonite buffer, it has a critical challenge due to its sensor size, which can disturb the overall integrity of the bentonite buffer during the initial installation process. In contrasts, the electrical resistivity test, widely known as a non-destructive method, is used to predict soil properties such as the degree of saturation and water contents. This method measures the electric resistance of materials using electric current induced by electric potential difference between two electrodes. Notably, there is no study that assess the integrity of bentonite buffer in a nuclear waste repository using electrical resistivity measurement. This study presents the electrical resistance numerical module under steady state using commercial finite element method (FEM), and quantitatively estimate the change of electrical resistance according to saturation and erosion of bentonite buffer. Furthermore, the electric potential and current density distribution formed between two electrodes are analyzed.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.

Post Irradiation Examination Facility (PIEF) is a test facility for nuclear fuel research and development and performance evaluation. From the past to the present, assemblies and fuel rods have been transported from nuclear power plants (NPP) several times, and various destructive and non-destructive tests have been performed. Among these, in the case of the 14×14 Westinghouse STD assemblies that are transported as a whole assembly, the top nozzle is connected to the guide tube by welding. Therefore, the fuel rods could not be removed from the assembly at the NPP, so the assemblies were transported to PIEF as is. Then, after cutting between the top nozzle and the guide tube in the pool, and the fuel rods were extracted and tested. In order to transport the assembly in the future, it is necessary to maintain stability by inserting the dummy rod into the unit cell from which the fuel rod is extracted. However, since the length of the dummy rod is almost 4 m and the diameter is about 10 mm, the dummy rod often bends while passing through the dimple spring of the grid. Additionally, when dummy rods are inserted into unit cells that are continuously empty after the fuel rods are extracted, there may be cases where the dummy rods are not inserted into the desired unit cell but are bent and incorrectly inserted into the next unit cell. The moment the dummy rods are inserted into the dimple spring of grid, a load is applied to the dummy rod due to the tension of the spring. If it can be inserted while offsetting the load, the work can be performed more smoothly. Accordingly, an underwater handling tool was developed that can be inserted while offsetting the tension of the spring. Using this handling tool applies a load to the dummy rod and rotates the dummy rod itself, offsetting the tension of the spring and allowing the dummy rod to be inserted without bending. This handling tool is equipped with a shock absorbing device to protect the dummy rod and spring, and a module to rotate the dummy rod. As a result of inserting the dummy rod using the developed handling tool, it was possible to easily insert the dummy rod into unit cells that were previously impossible to insert.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

Nuclear power is responsible for a large portion of electricity generation worldwide, and various studies are underway, including the design of permanent deep geological disposal facilities to safely isolate spent nuclear fuel generated as a result. However, through the gradual development of drilling technology, various disposal option concepts are being studied in addition to deep geological disposal, which is considered the safest in the world. So other efforts are also being made to reduce the disposal area and achieve economic feasibility, which requires procedures to appropriately match the waste forms generated from separation process of spent nuclear fuel with disposal option systems according to their characteristics. And safety issue of individual disposal options is performed through comparison of nuclide transport. This study briefly introduces the pre-disposal nuclide management process and waste forms, and also introduces the characteristics of potential disposal options other than deep geological disposal. And environmental conditions and possible pathways for nuclide migration are reviewed to establish transport scenarios for each disposal option. As such, under this comprehensive understanding, this study finally seeks to explore various management methods for high-level radioactive waste to reduce the environmental burden.

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

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

To address the pressing societal concern in Korea, characterized by the imminent saturation of spent nuclear fuel storage, this study was undertaken to validate the fundamental reprocessing process capable of substantially mitigating the accumulation of spent nuclear fuel. Reprocessing is divided into dry processing (pyro-processing) and wet reprocessing (PUREX). Within this context, the primary focus of this research is to elucidate the foundational principles of PUREX (Plutonium Uranium Redox Extraction). Specifically, the central objective is to elucidate the interaction between uranium (U) and plutonium (Pu) utilizing an organic phase consisting of tributyl phosphate (TBP) and dodecane. The objective was to comprehensively understand the role of HNO3 in the PUREX (Plutonium Uranium Redox Extraction) process by subjecting organic phases mixed with TBPdodecane to various HNO3 concentrations (0.1 M, 1.0 M, 5.0 M). Subsequently, the introduction of Strontium (Sr-85) and Europium (Eu-152) stock solutions was carried out to simulate the presence of fission products typically contented in the spent nuclear fuel. When the operation proceeds, the complex structure takes the following form. 􀜷􀜱􀬶 􀬶􀬾(􀜽􀝍) + 2􀜰􀜱􀬷 􀬿(􀜽􀝍) + 2􀜶􀜤􀜲(􀝋􀝎􀝃) ↔ 􀜷􀜱􀬶(􀜰􀜱􀬷)􀬶 ∙ 2􀜶􀜤􀜲(􀝋􀝎􀝃) Subsequently, separate samples were gathered from both the organic and aqueous phases for the quantification of gamma-rays and alpha particles. Alpha particle measurements were conducted utilizing the Liquid Scintillation Counter (LSC) system, while gamma-ray measurements were carried out using the High-Purity Germanium Detector (HPGe). The distribution ratio for U, Eu (Eu-152), and Sr (Sr-84) was ascertained by quantifying their activity through LSC and HPGe. Through the experiments conducted within this program, we have gained a comprehensive understanding of the selective solvent extraction of actinides. Specifically, uranium has been effectively separated from the aqueous phase into the organic phase using a combination of tributyl phosphate (TBP) and dodecane. Subsequently, samples containing U(VI), Eu(III), and Sr(II) underwent thorough analysis utilizing LSC and HPGe detectors. Our radiation measurements have firmly established that the concentration of nitric acid enhances the selective separation of uranium within the process.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

The radionuclide management process is a conditioning technology to reduce the burden of spent fuel management, and refers to a process that can separate and recover radionuclides having similar properties from spent fuels. In particular, through the radionuclide management process, high heat- emitting, high mobility, and high toxicity radionuclides, which have a significant impact on the performance of disposal system, are separated and managed. The performance of disposal system is closely related to properties (decay heat and radioactivity) of radioactive wastes from the radionuclide management process, and the properties are directly linked to the radionuclide separation ratio that determines the composition of radionuclides in waste flow. The Korea Atomic Energy Research Institute have derived process flow diagrams for six candidates for the radionuclide management process, weighing on feasibility among various process options that can be considered. In addition, the GoldSim model has been established to calculate the mass and properties of waste from each unit process of the radionuclides management process and to observe their time variations. In this study, the candidates for the radionuclide management process are evaluated based on the waste mass and properties by using the GoldSim model, and sensitivity analysis changing the separation ratio are performed. And the effect of changes in the separation ratio for highly sensitive radionuclides on waste management strategy is analyzed. In particular, the separation ratio for high heat-emitting radionuclides determines the period of long-term decay storage.

After the decision of the Wolsong unit 1 permanent shutdown (2019), spent fuel stored in the spent fuel bay (hereafter, SFB) should be transported to a dry storage facility (MACSTOR or Canister) in order to decommission Wolsong unit 1. Accordingly, KHNP has established a shipment schedule for damaged fuel of Wolsong Unit 1 and is trying to complete the shipment according to the schedule. Wolsong is equipped with transportation casks and dry storage facilities, but baskets need to be manufactured separately. In addition, license approval is required for baskets, transport cask, and dry storage facilities for legal grounds to contain, transport, and store damaged fuels. In this paper, the initial model, upgrade model, and automation model of encapsulation equipment planned to be introduced in Canada to handle PHWR’s damaged fuel were compared, and the optimal model was selected in consideration of KHNP’s planned spent fuel shipment schedule. The PHWR’s damaged fuel encapsulation system is a system developed the PHWR’s damaged spent fuel to be handled in the same way as the existing PHWR when storing it in the dry storage facility and loading a basket for capsulation into transport cask. At the Gentilly-2 nuclear power plant in Canada, a manually operated encapsulation system was used due to the low quantity of damaged fuel, which can be encapsulated two bundles a day, and this model is an initial model. In the case of Wolsong Unit 1, it has about 300 damaged fuels, so it takes about nine months to work when using the initial model. The upgrade model developed to improve work efficiency and reliability has increased work efficiency through some automation, but it would take about eight months to process the damaged spent fuels of Wolsong Unit 1, and this model has not yet been manufactured and applied. Lastly, the automation model changed the work location outside the SFB and automated drainage/drying operations. It is easy to maintain and replace consumables because the work is carried out by lifting the damaged fuel to a shuttle outside the SFB surrounded by a shielding chimney. Considering the reduction of drainage/drying time, it is possible to save about four times as much time as the initial model. That is, if the automation model is used, it is judged that the supply of Wolsong Unit 1 can be processed in about two months. However, in terms of license, initial model and upgrade model are expected to be easier and the period is expected to be shortened. However, if licensing is carried out as soon as equipment design is completed, it is believed that the period can be shortened by parallel equipment manufacturing and licensing. It is judged that the best way to comply with the target schedule is to select an automation model with excellent work performance, develop equipment, and proceed with licensing at the same time. Accordingly, KHNP is in the process of designing equipment with the aim of using the automation model to take out damaged fuel for Wolsong Unit 1.

On a global scale, the storage of spent nuclear fuel (SNF) within nuclear power plants (NPP) has become an important research topic due to limited space caused by approaching capacity saturation. SNF have e been collected over decades of NPP operation, coming up to capacity limitation. In case of Korea, every reactor except Saeul 1 and 2 has reached a SNF storage saturation rate of over 75%. One of the most studied methods for enhancing storage capacity efficiency involves increasing storage density using racks with neutron absorbers. Neutron absorbers like borated stainless steel (BSS) are utilized to manage the reactivity of densely stored SNF. However, major challenges of applying BSS are manufacturing hardness from heterogenous microstructure and mechanical property degradation from helium bubble formation. This study suggests that innovative fabrication methods of 3D printing can be good candidate for easier fabrication and better structural integrity of BSS. Directed energy deposition (DED), one of the 3D printing methods have become major candidate method for various alloys. It deposits alloy powder on base melt surface by high intensity laser, similar with welding process. Powder manufacturing is already demonstrated superior performance compared to casting in ASTM-A887, such as increased mechanical properties, owing to its well distributed chemistry of alloy. Moreover, as its original microstructural property, the formation of micro-pores through DED could lead to long-term performance improvements by capturing helium generated from the neutron absorption of boron. The potential for fabricating complex structure is also among the advantages of DED-produced neutron absorbers. Expected challenge on DED application on BSS is lack of printing condition data, because the 3D printing process have to be kept very careful variables of thermal intensity, powder flux and etc. These processes may get through much of trial & error for initial condition approaching. Nonetheless, as a recommendation of improved neutron absorber for efficient SNF pool storage, the concept of 3D printed BSS stands out as an intriguing avenue for research.

Once discharged, spent nuclear fuel undergoes an initial cooling process within deactivation pools situated at the reactor site. This cooling step is crucial for reducing the fuel’s temperature. Once the heat has sufficiently diminished, two viable options emerge: reprocessing or interim storage. A method known as PUREX, for aqueous nuclear reprocessing, involves a chemical procedure aimed at separating uranium and plutonium from the spent nuclear fuel. This separation not only minimizes waste volume but also facilitates the reuse of the extracted materials as fuel for nuclear reactors. The transformation of uranium oxides through dissolution in nitric acid followed by drying results in uranium taking the form of UO2(NO3)2 + 6H2O, which can then be converted into various solid-state configurations through different heat treatments. This study specifically focuses on investigating the phase transitions of artificially synthesized UO2(NO3)2 + 6H2O subjected to heat treatment at various temperatures (450, 500, 550, 600°C) using X-ray Diffraction (XRD) analysis. Heat treatments were also conducted on UO2 to analyze its phase transformations. Additionally, the study utilized XRD analysis on an unidentified oxidized uranium oxide, UO2+X, and employed lattice parameters and Bragg’s law to ascertain the oxidation state of the unknown sample. To synthesize UO2(NO3)2 + 6H2O, U3O8 powder is first dissolved in a 20% HNO3 solution. The solid UO2(NO3)2 + 6H2O is obtained after drying on a hotplate and is subsequently subjected to heat treatment at temperatures of 450, 500, 550, and 600°C. As the heat treatment temperature increases, the color of the samples transitions from orange to dark green, indicating the formation of different phases at different temperatures. XRD analysis confirms that uranyl nitrate, when heattreated at 500 and 550°C, oxidizes to UO3, while the sample subjected to 600°C heat treatment transforms into U3O8 due to the higher temperature. All samples exhibit sharp crystal peaks in their XRD spectra, except for the one heat-treated at 450°C. In the second experiment, the XRD spectra of the heat-treated UO2 consistently indicate the presence of U3O8 rather than UO3, regardless of the temperature. Under an oxidizing atmosphere within a temperature range of 300 to 700°C, UO2 can be oxidized to form U3O8. In the final experiment, the oxidation state of the unknown UO2+X was determined using Bragg’s law and lattice parameters, revealing that it was a material in which UO2 had been oxidized, resulting in an oxidation state of UO2.24.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

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.

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.

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.

To ensure the long-term supply and sustainability of uranium fuel, exploring alternative resources is essential, particularly considering that terrestrial reserves of uranium are limited (about 4.6 million tons). Since the amount of uranium dissolved in seawater is approximately 1000 times that of terrestrial reserves (i.e., about 4.5 billion tons), uranium extraction from seawater (UES) can be an alternative resource. However, the ultra-low concentration of uranium in seawater (about 3.3 ppb) poses a significant challenge in achieving economic feasibility for UES. This paper introduces case studies on the cost analysis of systems for recovering uranium from seawater, specifically focusing on braided fiber-based adsorbents developed by JAEA and ORNL. The cost analysis has been conducted based on using the deployment of these adsorbents on the bottom of the sea, which is a passive deployment method, thereby reducing the total costs of recovery. The analysis results can be used to identify R&D areas necessary for reducing cost components, making UES economically feasible.

An eco-friendly material was synthesized through interfacial polymerization of aniline on particles of g-C3N4 with arginine, resulting in Arg-PANI@g-C3N4 composite. The as-synthesized composite was characterized by the Brunauer, Emmett, and Teller (BET) surface area, X-ray energy dispersive spectroscopy (EDS), Fourier transform infrared spectroscopy (FTIR), scanning electron microscopy (SEM), and X-ray diffraction (XRD). The adsorption capability of as-synthesized composite towards Orange G (OG) dye has been evaluated under several experimental conditions, such as the adsorbent dosage, initial dye concentration, contact time under agitation, pH of dye solution and temperature. Thermodynamics parameters such as free energy (ΔG°), entropy (ΔS°), and enthalpy (ΔH°) were also calculated and suggested that the adsorption process is spontaneous and endothermic in nature. The kinetics data revealed that the adsorption of OG dye onto Arg-PANI@g-C3N4 follows the pseudo-second order kinetics model. The maximum adsorption capacity was found to be 80.54 mg·g−1. Furthermore, the Arg-PANI@g-C3N4 surface exhibited a Langmuir-like adsorption isotherm in contrast to a Freundlich isotherm due to homogeneous active site distribution. Regeneration investigation showed the excellent reusability of Arg-PANI@g-C3N4 composite during the cleaning up of solution containing OG dye molecules.