먹노린재 합성집합페로몬 후보물질 12종에 대한 유인력을 평가하였다. 페로몬 방출기는 4구 후각계 (Olfacomter)를 사용하였고 4개의 진공관에 각각 후보물질을 투입 후 진공 유압 방식으로 방출하여 포집기에 유인되는 먹노린재의 개채수를 측정하는 방법으로 검정하였다. 4구 후각계 페로몬 평가 방법은 기존 Y-관 후각 계의 문제점인 양방향 선택성과 공간 한정성을 개선하여 평가의 정확성을 향상시켰다. 유인력 평가 결과 12종의 유인제 후보물질 중 Trans-2-Decenal이 먹노린재 실험개체에 대하여 100% 유인력을 보였다. 또한, 선별된 Trans-2-Decenal의 먹노린재 유인력에 대한 유효농도 시험을 진행 한 결과, 50%의 농도에서 유인력이 가장 높았 다. 본 연구를 통해서 선별된 Trans-2 Decenal은 기존의 노린재과에 대한 페로몬 트랩에 비해 먹노린재에 대한 유인 효과가 높을 것으로 사료되었다. 이에 따라, Trans-2 Decenal을 기반으로 한 페로몬 트랩이 상용화된다면 추후의 먹노린재 방제 효과가 높아질 것으로 기대된다.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

Molten Salt Reactor (MSR) is one of the 4th generation nuclear power systems which is its verified technology in physically and chemically. Among the various salts used for MSR system, the eutectic composition of NaCl-MgCl2 system maintains the liquid state at around 450°C, in the same time, it has high solubility for nuclear fuel chlorides. This characteristic has high advantage for lowering the operating temperature for the MSR, which could reduce the problem of hightemperature corrosion by salt for structural materials significantly. In particular, since MgCl2 has the similar standard reduction potential with nuclear fuel, is used as a surrogate for, many basic researches have been conducted for verifying characteristic of MgCl2. It is well-known that main short-advantage of MgCl2 is hygroscopic properties. MgCl2 changes to MgCl2-xH2O state easily by absorbing moisture in air condition. The hydrated MgCl2 is producing MgOHCl by thermally decomposing at high temperature, the formed MgOHCl corrodes structural materials, even small amount of MgOHCl gives significant damage. Therefore, the purification of MgCl2 has been required for long-term operation of MSR using MgCl2 as a base salt. In this study, the purification of eutectic composition salt for NaCl-MgCl2 has been mainly performed by considering its thermodynamic properties and electrochemical characteristic, and the experimental results have been discussed.

The presence of organic components in spent scintillation liquid should be considered during all steps of radioactive waste processing for final disposal. Scintillation liquids often referred to as cocktails are generated form radiochemical analyses of radionuclides, which mainly consists of mixtures of liquid organic materials such as toluene and xylene. Typical features of these liquid organic materials are volatility, combustibility and toxicity. These are the reason why special attention must be paid to the management of liquid organic radioactive wastes. To select an appropriate waste management strategy and to design the treatment process of spent scintillation cocktails, it is required to investigate the nature of the waste such as specific radioactivity and moisture content. The analysis results of spent scintillation liquid generated at Wolsong nuclear power plants will be discussed. An overview of the technical approaches available for the treatment of organic radioactive waste will be additionally provided.

In KAERI, the nuclide management technology is currently being developed for the reduction of disposal area required for spent fuel management. Among the all fission products of interest, Cs, I, Kr, Tc are considered to be significantly removed by following mid-temperature and hightemperature treatment, however, a difficulty of real spent-fuel thermal treatment experiment limits the development of such thermal treatment. The test employing SimFuel (Simulated Spent Fuel) can be an alternative for such condition, however, the fabrication of SimFuel containing semivolatile species such as Cs, I and Re (substitute for Tc) was not achieved for conventional sintering method since such species are easily removed during hot temperature treatment. In this study, for the prevention of volatilization of such species and the inclusion of semi-volatile species in fabrication of SimFuel, argon-based high pressurizing up to Max 100 bar was considered to be applied in high temperature treatment. For this, lab-scale hot-isostatic press applicable up to 1,500°C was fabricated and is being waiting for the approval for high-pressure test. After approval of license, UO2 baesd SimFuel containing CsI will be fabricated and its micro-structure and composition will be evaluated through SEM-EDX and XRD

In KAERI’s previous phosphate precipitation tests, the dispersed powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with various metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides composed of actinides such as uranium and three rare earths (Nd, Ce and La) with lithium phosphate is a solid-liquid reaction. A phosphorylation reaction rate is very fast and the metal phosphates as a reaction product precipitated on the bottom of the molten salt crucible. One of the recovery methods of the metal phosphate precipitates is segregation the lower part (precipitates) of the salt ingot using the various cutting tools. Recently, a new phosphorylation experiment using lithium phosphate ingots carried out in order to collect the metal phosphate precipitates into a small recovering vessel, and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is extremely slower than that of test using lithium phosphate powder. In this study, the precipitation reactor design (a tapered crucible with polished inner surface) used for phosphorylation reaction showed that the salt ingot with metal phosphate precipitates could be detached from a tapered stainless steel crucible. We propose that the recovery of precipitates from a salt ingot is possible by introducing a dividing plate structure into a molten salt and by positioning it at the interface between salt and precipitated metal phosphate.

Currently, the Korea Atomic Energy Research Institute is conducting research on the development of technology to reduce the disposal area for SF (Spent nuclear Fuel). If the main radionuclides contained in SF can be separated and recovered according to their characteristics (long half-life, high mobility and high heat load) and uranium oxide which is expected to be the final residue, can be made into solids, the burden of the permanent disposal area of the SF will be greatly reduced. The waste form that end up in the repository must be verified for ease of manufacture and stability of the block. And, in order to increase the loading efficiency, a large block manufacturing technology is needed. This study describes the background of introducing PSA (Particle Size Analyzer) which is one of the necessary equipment for manufacturing UO2 blocks using slip casting, the method of using the equipment and performance verification of the equipment using standard samples. The particle size affects the sintering quality by the way the particles rearrange themselves during sintering. Powders of small particles are generally less free flowing and more difficult to compress, they form thin pores between the particles and sinter to higher density. In contrast, larger particle has a lower sintered density. Therefore, accurate particle size measurement and the selection of a suitable particle size are important. For this purpose, a PSA was installed in nuclear cycle experiment research center. To verify the performance of the equipment, a standard sample of 1.025 μm was analyzed. We got an average particle size of 1.0293 μm and standard deviation of 0.0668 μm. This value was within the uncertainty(±0.018 μm) of the sample’s certificate. In the future, this equipment will measure the size of UO2 (depleted uranium) powder and to produce large scale uranium oxide blocks.

Pyroprocessing technology has emerged as a viable alternative for the treatment of metal/oxide used fuel within the nuclear fuel cycle. This innovative approach involves an oxide reduction process wherein spent fuel in oxide form is placed within a cathode basket immersed in a molten LiCl-Li2O salt operating at 923 K. The chemical reduction of these oxide materials into their metallic counterparts occurs through a reaction with Li metal, which is electrochemically deposited onto the cathode. However, during process, the generation of Li2O within the fuel basket is inevitable, and due to the limited reduction efficiency, a significant portion of rare earth oxides (REOx) remains in their oxide state. The presence of these impurities, specifically Li2O and REOx, necessitates their transfer into the electrorefining system, leading to several challenges. Both Li2O and REOx exhibit reactivity with UCl3, the primary electrolyte within the electrorefining system, causing a continuous reduction in UCl3 concentration throughout the process. Furthermore, the formation of fine UO2 powder within the salt system, resulting from chemical reactions, poses a potential long-term operational and safety concern within the electrorefining process.Various techniques have been developed to address the issue of UO2 fine particle removal from the salt, utilizing both chemical and mechanical methods. However, it is crucial that these methods do not interfere with the core pyroprocessing procedure. This study aims to investigate the impact of Li2O and REOx introduced from the electrolytic reduction process on the electrorefining system. Additionally, we propose a method to effectively eliminate the generated UO2 fine powder, thereby enhancing the long-term operational stability of the electrorefining process. The efficiency of this proposed solution in removing oxidized powder has been confirmed through laboratory-scale testing, and we will provide a comprehensive discussion of the detailed results.

It is known that ZrCl4 can be used in the chlorination process of spent nuclear fuel. However, its solubility in high temperature molten salt is very small, making it difficult to dissolve a large amount of ZrCl4. Therefore, in this study, a flange-type sealed reactor was manufactured to observe the reaction characteristics of LiCl-KCl salt and ZrCl4. LiCl-KCl salt and ZrCl4 were placed in each alumina crucible, the reactor was sealed, and heated. The temperature at the reactor surface was above 500°C and maintained at that temperature for 48 hours. After completion of the reaction, the reactor was opened and the reaction products were recovered from each alumina crucible. The crystal structure of the reaction product was identified through XRD analysis, and the concentration of Zr was analyzed using ICP. Reaction characteristics were observed according to the molar ratio of ZrCl4 added to the number of moles of KCl in LiCl-KCl salt. The molar ratios of ZrCl4 to KCl were 0.5, 1, 2, and 3, respectively. As a result of each experiment, more than 95% of the injected ZrCl4 was vaporized and there was almost no residue in the ZrCl4 crucible. In the LiCl- KCl crucible, the weight increased in proportion to the amount of ZrCl4 added. As a result of XRD analysis, K2ZrCl6 was confirmed in all LiCl-KCl salt product. When the ZrCl4/KCl molar ratio was 2 and 3, LiCl-KCl could not be confirmed. Additionally, when the ZrCl4/KCl molar ratio was 1, LiCl was identified, but KCl was not found. Almost all of the KCl appears to have reacted with ZrCl4. ICP analysis results showed that the Zr concentration was proportional to the amount of ZrCl4 added in each LiCl-KCl salt, and exceeding the number of moles of reaction with KCl in the LiCl-KCl salt was observed. Therefore, these experimental results showed that ZrCl4 can be dissolved in LiCl-KCl salt at a maximum concentration higher than its solubility.

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.

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.

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.

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.

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.

먹노린재의 방제시기 설정을 위하여 서식지 내 발생 양상을 조사하였다. 전라남도를 중심으로 돌발 대량 발생 한 먹노린재의 발생 조사는 2023년도 벼의 모내기가 완료된 시점부터 전남 곡성군 석곡면과 여수시 화양면 일대 의 친환경단지에서 주 1회 실시하였으며, 벼의 수확이 완료되는 시점까지 진행될 예정이다. 금년도 발생 조사 결과는 지난 2021년도 결과와 비교하여 분석하고자 하였다. 현제까지의 결과, 여수시의 친환경단지 내 먹노린재 의 발생은 모내기가 완료된 이 후 2주 경과 시점에서 발견되었으며, 곡성군은 4주 경과 시점에서 발견되었고, 발생 최고점은 여수시의 경우 모내기 후 6주 경과 시점으로, 곡성군의 경우 5주 경과 시점으로 나타났다. 먹노린재 의 대발생이 지속되던 2021년도의 결과와 비교하였을 때, 2023년도 결과도 유사한 것으로 나타났다. 이에 따라, 먹노린재 친환경제재 살포 시기는 모내기 후 2주 경과 시점이 적절한 것으로 나타났으며, 2차 방제 시점에 대해서 는 추가 조사를 통해서 제안하고자 한다.

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl2, NH4Cl, and Cl2 as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl2, NH4Cl, and Cl2). Following heat treatment, the utilization of MgCl2 as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, 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 the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

The disposal of spent nuclear fuel (SNF) poses a significant challenge due to its high radioactivity and heat generation. However, SNF contains reusable materials, such as uranium and trans-uranium, which can be recovered through aqueous reprocessing or pyrochemical processes. Prior to these processes, voloxidation is necessary to increase reaction kinetics by separating fuels from cladding and reducing the particle size. In the voloxidation, uranium dioxide (UO2) from SNF is heated in the presence of oxygen and oxidized to triuranium octoxide (U3O8), resulting a release of gaseous fission products (FPs), including technetium-99 (Tc-99), which poses a risk to human health and the environment due to its high mobility and long half-life of 2.1×105. To date, various methods have been developed to capture Tc in aqueous solutions. However, a means to capture the gaseous form of Tc (Tc2O7) is essential in the voloxidation. Due to the radioactive properties of technetium isotopes, rhenium is often used as a substitute in laboratory settings. The chemical properties of rhenium and technetium, such as their electronic configurations, oxidation states, and atomic radii, are similar and these similarities indicates that the adsorption mechanism for rhenium can be analogous to that for technetium. In the previous study, a disk-type adsorbent based on CaO developed was effective in capturing Re. However, this study lacked sufficient data on the chemical properties and capture performance of the adsorbent. Furthermore, the fabrication of disk-type adsorbents is time-consuming and requires multiple steps, making it impractical for mass production. This study introduces a simple and practical method for preparing CaO-based pellets, which can be used as an adsorbent to capture Re. The results provide a better understanding of the adsorption behavior of CaO-based pellets and their potential for capturing Tc-99. To the best of our knowledge, this is the first study to apply a CaO-based pellet to capture Re and investigate its potential for capturing Tc-99.