Selective Chlorination of Actinide and Alkali Earth Elements in Spent LWR Nuclear Fuel Using Ammonium Chloride

Dalsung Yoon,Ju ho Lee,Seungwoo Paek,Sang-Kwon Lee,Taeho Kim,Chang Hwa Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1

As temporary storage facilities for spent nuclear fuel (SNF) are becoming saturated, there is a growing interest in finding solutions for treating SNF, which is recognized as an urgent task. Although direct disposal is a common method for handling SNF, it results in the entire fuel assembly being classified as high-level waste, which increases the burden of disposal. Therefore, it is necessary to develop SNF treatment technologies that can minimize the disposal burden while improving long-term storage safety, and this requires continuous efforts from a national policy perspective. In this context, this study focused on reducing the volume of high-level waste from light water reactor fuel by separating uranium, which represents the majority of SNF. We confirmed the chlorination characteristics of uranium (U), rare earth (RE), and strontium (Sr) oxides with ammonium chloride (NH4Cl) in previous study. Therefore, we prepared U-RE-SrOx simulated fuel by pelletizing each elements which was sintered at high temperature. The sintered fuel was again powdered by heating under air environment. The powdered fuel was reacted with NH4Cl to selectively chlorinate the RE and Sr elements for the separation. We will share and discuss the detailed results of our study.

A Study on Optimization of UCl3 Production by Reaction of Uranium Metal and Ammonium Chloride

Dalsung Yoon,Seungwoo Paek,Sang-Kwon Lee,Ju Ho Lee,Taeho Kim,Chang Hwa Lee 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2

Interests in molten salt reactor (MSR) using a fast spectrum (FS) have been increased not only for having a high power density but for burning the high-level waste generated from nuclear power plants. For developing the FS-MSR technologies, chloride-based fuels are considered due to the advantage of higher solubility of actinides and lanthanides over fluoride-based salts. Despite significant progress in development of MSR technology, the manufacturing technology for production of the fuel is still insufficiently understood. One of the option to prepare the MSR fuel is to use products from pyroprocessing where oxide form of spent nuclear fuel is reduced into metal form and useful elements can be collected via electrochemical methods in molten salt system at high temperature. In order to chlorinate the products into chloride form, previous study used NH4Cl to chlorinate U metal into UCl3 in an airtight reactor. It was found that the U metal was completely chlorinated into chloride forms; however, impurities generated by the reaction of NH4Cl and reactor wall were found in the product. Therefore, in this work, the air tight reactor was re-deigned to avoid the reaction of reactor wall by insertion of Al2O3 crucible inside of the reactor. In addition, the reactor size was increased to produce UCl3 over 100 g. Using the newly designed reactor, U metal chlorination experiments using NH4Cl chlorinating agent were performed to confirm the optimal experimental conditions. The detailed results will be further discussed.

Chlorination of Simulated Oxide Fuel to Minimize Volume of High Level Nuclear Fuel Waste

Dalsung Yoon,Seungwoo Paek,Changhwa Lee 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

Facing the problem of saturation of spent nuclear fuel (SNF) stored in temporary storage facilities on sites, interest in the treatment of SNF is increasing, and it is recognized as a task that needs to be solved promptly. Although direct disposal is a general method for dealing with SNF, the entire fuel assembly is classified as high-level waste; thus, the burden of disposal is high. In order to minimize the disposal burden with enhancing safety for long term storage, it is necessary to develop SNF treatment technologies and continuous efforts are required from a national policy perspective. The present study focused on minimizing the volume of high level waste from light water reactor fuel by separation of uranium, which accounts for most of SNF. The chlorination characteristics of uranium (U), rare earth (RE) oxides were confirmed through lab-scale experiments, and the possibility of uranium separation from U-RE simulated fuel was evaluated using NH4Cl chlorinating agent. The detailed results will be posted and discussed.

Uranium Metal Chlorination Using NH4Cl for the Preparation of MSR Fuel

Dalsung Yoon,Seungwoo Paek,Sangkwon Lee,Ju ho Lee,Changhwa Lee 한국방사성폐기물학회 2021 한국방사성폐기물학회 학술논문요약집 Vol.2021 No.11

Fundamental Study on Uranium Oxide Separation From the Spent Nuclear Fuel by Chlorination Using NH4Cl

Dalsung Yoon,Seungwoo Paek,Ju Ho Lee,Changhwa Lee 한국방사성폐기물학회 2021 한국방사성폐기물학회 학술논문요약집 Vol.2021 No.11

Actinide Drawdown From LiCl-KCl Eutectic Salt via Galvanic/chemical Reactions Using Rare Earth Metals

Yoon, Dalsung,Paek, Seungwoo,Jang, Jun-Hyuk,Shim, Joonbo,Lee, Sung-Jai Korean Radioactive Waste Society 2020 방사성폐기물학회지 Vol.18 No.3

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

Enhancing Operational Stability in Pyroprocessing: Study on UO2 Powder Removal in Electrorefining Process

Dalsung Yoon,Sang-Kwon Lee,Juho Lee,Seungwoo Paek,Taeho Kim,Chang Hwa Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

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.

Activin A/BMP2 Chimera (AB204) Exhibits Better Spinal Bone Fusion Properties than rhBMP2

Ryu, Dalsung,Yoon, Byung-Hak,Oh, Chang-Hyun,Kim, Moon-Hang,Kim, Ji-Yong,Yoon, Seung Hwan,Choe, Senyon The Korean Neurosurgical Society 2018 Journal of Korean neurosurgical society Vol.61 No.6

Objective : To compare the spinal bone fusion properties of activin A/BMP2 chimera (AB204) with recombinant human bone morphogenetic protein (rhBMP2) using a rat posterolateral spinal fusion model. Methods : The study was designed to compare the effects and property at different dosages of AB204 and rhBMP2 on spinal bone fusion. Sixty-one male Sprague-Dawley rats underwent posterolateral lumbar spinal fusion using one of nine treatments during the study, that is, sham; osteon only; $3.0{\mu}g$, $6.0{\mu}g$, or $10.0{\mu}g$ of rhBMP2 with osteon; and $1.0{\mu}g$, $3.0{\mu}g$, $6.0{\mu}g$, or $10.0{\mu}g$ of AB204 with osteon. The effects and property on spinal bone fusion was calculated at 4 and 8 weeks after treatment using the scores of physical palpation, simple radiograph, micro-computed tomography, and immunohistochemistry. Results : Bone fusion scores were significantly higher for $10.0{\mu}g$ AB204 and $10.0{\mu}g$ rhBMP2 than for osteon only or $1.0{\mu}g$ AB204. AB204 exhibited more prolonged osteoblastic activity than rhBMP2. Bone fusion properties of AB204 were similar with the properties of rhBMP2 at doses of 6.0 and $10.0{\mu}g$, but, the properties of AB204 at doses of $3.0{\mu}g$ exhibited better than the properties of rhBMP2 at doses of $3.0{\mu}g$. Conclusion : AB204 chimeras could to be more potent for treating spinal bone fusion than rhBMP2 substitutes with increased osteoblastic activity for over a longer period.

Reaction of Rare Earth Oxide From the Simfuel and Chlorination Reagents in LiCl-KCl Molten salt

Seungwoo Paek,Dalsung Yoon,Ju Ho Lee,Sang Kwon Lee,Chang Hwa Lee 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

Reaction of ZrCl4 and LiCl-KCl Molten Salt

Seungwoo Paek,Dalsung Yoon,Sang Kwon Lee,Chang Hwa Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

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.