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.

Nuclear safety, security, and safeguards (nuclear 3S) are essential components for establishing robust nuclear environments. Nuclear safety is to protect public and environments from radioactive contamination, which can be caused in accidents. Nuclear security is to protect nuclear facilities from terrorism or sabotage, which related to physical a ttacks or insider threats. And nuclear safeguards is to protect nuclear materials from extortion by a state with a purpose of weaponizing activities. When a new nuclear facility is introduced, it is possible to save abundant amount of resources by considering nuclear 3S in an early stage (design phases). Initially, the international atomic energy agency (IAEA) recommended safeguards-by-design (SBD) approach. The concept of SBD gradually expands to nuclear 3S-by-design (3SBD). Though there are differences in purpose and target subject, each nuclear ‘S’ is closely related with others. When introducing a certain technology or equipment in order to enhance one ‘S’, another ‘S’ also get affected. The effect can be synergies or conflicts. For instance, confidential information in nuclear security is required for a safeguards activity. On the contrary, inspection equipment for safeguards can be used for security. Pyroprocessing is a technology for managing used nuclear fuels. As pyroprocessing is a backend fuel cycle technology, a sensitive nuclear technology, safeguards has taken a large portion of nuclear 3S research in an effort to achieve international credibility and nuclear transparency. As mentioned, there are both synergies and conflicts in integrating nuclear 3S. In this study, we investigate potential challenges in applying nuclear 3S integration to pyroprocessing by addressing synergies and conflicts. This approach will suggest required supplementary methods to build the reliable pyroprocessing environment.

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.

Since 1997, the Republic of Korea (ROK) has been developing pyro-processing (Pyro) technology to reduce the disposal burden of high-level radioactive waste by recycling spent nuclear fuel (SNF). Compared to plutonium and uranium extraction process, Korean Pyro technology has relatively excellent proliferation resistance that cannot separate pure plutonium owing to its intrinsic characteristics. Regarding Pyro technology development of ROK, the Bush administration considered that Pyro is not reprocessing under the Global Nuclear Energy Partnership, whereas the Obama administration considered that Pyro is subject to reprocessing. However, the Bush and Obama administrations did not allow ROK to conduct full Pyro activities using SNF, even though ROK had faithfully complied with international nonproliferation obligations. This is because the US nuclear nonproliferation policy to prevent the spread of sensitive technologies, such as enrichment and reprocessing, has a strong effect on ROK, unlike Japan, on a bilateral level beyond the NPT regime for non-proliferation of nuclear weapons.

Pyroprocessing is a promising technique for the treatment of damaged fuel debris (corium) generated by severe nuclear accidents. The debris typically consists of (U, Zr)O2 originating from the UO2 fuel and Zr alloy-based cladding. By converting the corium to a metallic form, the principal components of the fuel can be recovered through subsequent electrorefining, allowing for long-term storage or final disposal. A study investigated the reduction of zirconium oxide compounds by Li metal as a reductant in molten LiCl salt. This research explored the feasibility of treating damaged nuclear fuel debris, which mainly consists of (U, Zr)O2. The results showed that ZrO2 was successfully reduced to Zr metal by Li metal in LiCl salt at 650C without the formation of Li2ZrO3. In particular, Zr metal was produced without the formation of Li2ZrO3 when LiCl salt containing a high concentration of Li metal was used. However, Zr metal was produced with Li2ZrO3 when LiCl salt containing both Li metal and Li2O was added. This suggests that the concentration of Li metal in the LiCl salt is an important factor in determining the formation of Li2ZrO3. The study also demonstrated that Li2ZrO3 was partially reduced to Zr metal by Li metal in LiCl salt. This finding suggests that Li metal may be effective in reducing other oxide compounds in molten LiCl salt, which could be useful in the treatment of corium. Overall, the research provides valuable insights into the feasibility of using pyroprocessing for the treatment of corium. The ability to recover and store the principal components of the fuel through electrorefining could have important implications for the long-term management of nuclear waste.

Measuring the concentration of corrosion products or nuclear fission products (FPs) in molten salts is crucial for pyroprocessing and molten salt reactors. Electrochemical analysis methods that can be performed in situ offer significant advantages for monitoring the concentration of corrosion products or FPs in molten salts. A microelectrode is an electrode with a length of several tens of micrometers on one side. The use of a microelectrode for electrochemical analysis has several advantages due to its small size, including rapidly reaching the limiting current regardless of the scan rate, immediate attainment of the limiting current upon applying an overpotential for instant monitoring within milliseconds, accurate measurement even in low convection systems, a small iR drop resulting from low flowing current and high signal accuracy, and high current density resulting in a high signal-tonoise ratio (SNR). Among various methods for making microelectrodes, techniques involving cutting a thin wire or using capillaries (such as the dual-bore capillary and pulled glass capillary methods) require precise manual skills and experience. Therefore, the results may vary depending on the maker’s skill level, and it can be difficult to control the electrode’s area, thickness, and surface uniformly. Recent research has focused on using semiconductor processes to fabricate microelectrodes, where CVD, metal sputtering, photolithography, and etching processes work together to deposit, refine, and shape the required material on a silicon wafer to create microelectrodes. However, the durability of microelectrodes produced this way is still low (usable for about 15-30 minutes), and there is no clear research on the degradation mechanism over time. To verify the proper operation of the fabricated microelectrodes, cyclic voltammetry (CV) is performed at various scan rates (from 10 mVs-1 to 2 Vs-1), and chronoamperometry (CA) is also examined to confirm whether the electrodes rapidly reach a steady-state current. After confirming their proper operation, CV is continuously measured until the microelectrodes are destroyed in a LiCl-KCl solution containing a small amount of FPs (Sm 340 mM) at 450°C. By observing changes in the electrical signal of the microelectrodes over time, the durability is evaluated, and the mechanism of performance degradation of the electrode is discovered. The experiment is then repeated by gradually increasing the temperature by 30°C from 450°C up to 600°C to observe the changes with temperature. This study provides basic information for future microelectrode experiments, and by diagnosing the cause of destruction, a more durable microelectrode structure can be manufactured.

Radioactive wastes, including used nuclear fuel and decommissioning wastes, have been treated using molten salts. Electrochemical sensors are one of the options for in-situ process monitoring using molten salts. However, in order to use electrochemical sensors in molten salt, the surface area must be known. This is because the surface area affects the current of the electrode. Previous studies have used a variety of methods to determine the electrode surface area in molten salts. One method of calculating the electrode surface area is to use the reduction current peak difference between electrodes with known length differences. The method is based on the reduction peak and has the benefit of providing long-term in-situ monitoring of surfaces immersed in molten salt. A number of assumptions have been made regarding this method, including that there is no mass transport by migration or convection; the reaction is reversible and limited by diffusion; the chemical activity of the deposit should be unity; and species should follow linear diffusion. For the purpose of overcoming these limitations, a variety of machine learning algorithms were applied to different voltammogram datasets in order to calculate the surface area. Voltammogram datasets were collected from multiarray electrodes, comprising a multiarray holder, two tungsten rods (1 mm diameter) working electrodes, a quasi-reference electrode, and a counter electrode. The multiarray electrode holder was connected to the auto vertical translator, which uses a servo motor, for changing the height of the rod in the molten salts. To make big and diverse data for training machine learning models, various concentrations of corrosion products (Cr, Fe) and fission products (Eu, Sm) in NaCl-MgCl2 eutectic salts were used as electrolyte; electrolyte temperatures were 500, 525, 550, 575, and 600°C. This study will demonstrate the potential of utilizing machine learning based electrochemical in situ monitoring in molten salt processing.

Spin-off pyroprocessing technology and inert anode materials to replace the conventional carbon-based smelting process for critical materials were introduced. Efforts to select inert anode materials through numerical analysis and selected experimental results were devised for the high-throughput reduction of oxide feedstocks. The electrochemical properties of the inert anode material were evaluated, and stable electrolysis behavior and CaCu generation were observed during molten salt recycling. Thereafter, CuTi was prepared by reacting rutile (TiO2) with CaCu in a Ti crucible. The formation of CuTi was confirmed when the concentration of CaO in the molten salt was controlled at 7.5mol%. A laboratory-scale electrorefining study was conducted using CuTi(Zr, Hf) alloys as the anodes, with a Ti electrodeposit conforming to the ASTM B299 standard recovered using a pilot-scale electrorefining device.

During electrorefining, fission products, such as Sr and Cs, accumulate in a eutectic LiCl-KCl molten salt and degrade the efficiency of the separation process by generating high heat and decreasing uranium capture. Thus, the removal of the fission products from the molten salt bath is essential for reusing the bath, thereby reducing the additional nuclear waste. While many studies focus on techniques for selective separation of fission products, there are few studies on processing monitoring of those techniques. In-situ monitoring can be used to evaluate separation techniques and determine the integrity of the bath. In this study, laser-induced breakdown spectroscopy (LIBS) was selected as the monitoring technique to measure concentrations of Sr and Cs in 550°C LiCl-KCl molten salt. A laser spectroscopic setup for analyzing high-temperature molten salts in an inert atmosphere was established by coupling an optical path with a glove box. An air blower was installed between the sample and lenses to avoid liquid splashes on surrounding optical products caused by laser-liquid interaction. Before LIBS measurements, experimental parameters such as laser pulse energy, delay time, and gate width were optimized for each element to get the highest signal-to-noise ratio of characteristic elemental peaks. LIBS spectra were recorded with the optimized conditions from LiCl-KCl samples, including individual elements in a wide concentration range. Then, the limit of detections (LODs) for Sr and Cs were calculated using calibration curves, which have high linearity with low errors. In addition to the univariate analysis, partial least-squares regression (PLSR) was employed on the data plots to obtain calibration models for better quantitative analysis. The developed models show high performances with the regression coefficient R2 close to one and root-mean-square error close to zero. After the individual element analysis, the same process was performed on samples where Sr and Cs were dissolved in molten salt simultaneously. The results also show low-ppm LODs and an excellent fitted regression model. This study illustrates the feasibility of applying LIBS to process monitoring in pyroprocessing to minimize nuclear waste. Furthermore, this high-sensitive spectroscopic system is expected to be used for coolant monitoring in advanced reactors such as molten salt reactors.

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.