Korean innovative SMR has been implemented developing with improved safety/economy and i- SMR technology development project to secure a competitive edge in SMR. For nuclear power plants, according to the revision of the Nuclear Safety Act (2013.6), it is mandatory to be reflected in the aging management program of nuclear power plants, and the aging management and regulation of major nuclear power plants are being strengthened. For i-SMR, chemistry environment and management strategy is essential to mitigate corrosion and radiation fields, since it has compacted and integrated module designs. Since 1994, zinc injection into the reactor coolant system (RCS) has been applied more than 100 PWRs in the world to mitigate primary water stress corrosion cracking (PWSCC) and to reduce outof- core radiation fields. In domestic NPPs, 7 have been applying zinc injection and had up to 90% radiation field reductions. For this reason, SMR needs to apply zinc injection for chemistry strategy. Zinc target concentration will be 5~40 ppb at i-SMR, based on Ni-Fe-Cr materials as same as PWRs. Zinc injection location is in volume and purification control system between the volume control tank and charging P/P where the pressure is moderate. Zinc injection skid can consist of two micro-controllable pump (one for operation and one for stand-by) and one injection tank (batching tank for zinc solution). Zn, Ni, Si, Fe, and activated corrosion products should be monitored to identify zinc injection controls and trends. Flux mapping for core performance monitoring should be evaluated. The application of zinc will be essential and effective and bring sustainable reliability for corrosion control and mitigation strategy to meet the risk-free i-SMR development.

The organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA) can enhance the radionuclides’ solubility and have the potential to induce the acceleration of radionuclides’ mobility to a far-field from the radioactive waste repository. Hence, it is essential to evaluate the effect of organic complexing agents on radionuclide solubility through experimental analysis under similar conditions to those at the radioactive waste disposal site. In this study, five radionuclides (cesium, cobalt, strontium, iodine, and uranium) and three organic complexing agents (EDTA, NTA, and ISA) were selected as model substances. To simulate environmental conditions, the groundwater was collected near the repository and applied for solubility experiments. The solubility experiments were carried out under various ranges of pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and concentrations of organic complexing agents (0, 10-5, 10-4, 10-3, and 10-2 M). Experimental results showed that the presence of organic complexing agents significantly increased the solubility of the radionuclides. Cobalt and strontium had high solubility enhancement factors, even at low concentrations of organic complexing agents. We also developed a support vector machine (SVM) model using some of the experimental data and validated it using the rest of the solubility data. The root mean square error (RMSE) in the training and validation sets was 0.012 and 0.016, respectively. The SVM model allowed us to estimate the solubility value under untested conditions (e.g., pH 12, temperature 30°C, ISA 5×10-4 M). Therefore, our experimental solubility data and the SVM model can be used to predict radionuclide solubility and solubility enhancement by organic complexing agents under various conditions.

The bioreduction process from soluble U(VI) to insoluble U(IV) has been extensively studied in the field of radionuclides migration. Since acetic acid (AcOH) is widely used as an electron donor for bioreduction of U(VI), it is necessary to understand the effect of U(VI)-AcOH complexes that exist in different species depending on pH on this process. Changes in samples before and after bioreduction can be compared using time-resolved laser luminescence spectroscopy (TRLLS), which measures the characteristic luminescence spectra of different U(VI) species. Although luminescence properties of U(VI)-AcOH species were reported, experiments were conducted under conditions below pH 4.5. In this study, spectrophotometry and TRLLS for U(VI)-AcOH species (10−100 μM U(VI) and 20 mM AcOH) were performed in pH ranges extending to neutral and alkaline pH regions similar to groundwater conditions as well as acidic pH region. Two different complexes (UO2(AcO)+, UO2(AcO)2 with U(VI) and acetate ratios of 1:1, 1:2) were observed in the acidic pH region. The 1:1 complex, which appears as the pH increases, has no luminescence properties, but its presence can be confirmed because it serves to reduce the luminescence intensity of UO2 2+. In contrast, the 1:2 complex exhibits distinct luminescence properties that distinguish it from UO2 2+. The 1:3 complex (UO2(AcO)3 -) expected to appear with increasing pH was not observed. Two different complexes ((UO2)3(OH)5 +, (UO2)3(OH)7 - with U(VI) and OH ratios of 3:5, 3:7) were observed as the major species in the neutral pH region, but their luminescence lifetimes are remarkably short compared those in the absence of AcOH. Solid U(VI) particles were observed in the alkaline pH region, and they also had completely different luminescence properties from the aforementioned U(VI)-AcOH and U(VI)-hydrolysis species. Based on these results, the effect of pH in the presence of AcOH on the bioreduction process from U(VI) to U(IV) will be discussed.

Noble metal precipitates are fission products that can be found as metallic alloys in used nuclear fuel. They do not exist homogenously inside the fuel pellets, but exists in grain boundaries in the form of immiscible particles. The first drawback that comes because they exist in grain boundaries is the degradation of mechanical integrity. The particles in the grain boundaries can be considered as defect n solid solution of uranium oxide pellets, and they can change the lattice volume. Therefore, it is known that it can cause stress corrosion cracking of fuel pellets. Furthermore, there is a negative effect from the perspective of used fuel management. However, they also have a positive effect on used fuel management. Since the noble metal has galvanic reduction effect, the particles serve as an oxidation inhibitor for uranium. There are many other effects regarding to the noble metal precipitates. However, in any case, quantifying the particles is important in order to quantitatively analyze these effects from the perspective of used fuel management. SCALE/TRITON code was applied to calculate the noble metal isotopes including Mo, Tc, Ru, Rh and Pd. In order to calculate the distribution inside the pin, the multiregion cell model was selected. In particular, a cylindrical geometry was used, and the pellet was divided into several layers. In addition, coolant and cladding surrounded the pellet. Finally, the radial distribution was evaluated using the computational code, along with neutron flux map.

Korea Atomic Energy Research Institute (KAERI) is planning to disposal of the radioactive contaminated cement waste form to the final disposal facility. The final disposal facility require evaluation of immersion, compressive strength, and radionuclide inventory of radioactive wastes to meet the acceptance criteria for safe disposal. According to the LILW acceptance criteria of the Nuclear Safety and Security Commission ok Korea (NSSC), the disposal limit radioactivity of 129I (3.70×101 Bq/g) is lower than other radionuclides. 129I emits low energy as its disposal limit is low, so it is difficult to analyze in the presence of 137Cs and 60Co which emit high energy. Therefore, it is essential to an accurately separate and analyze iodine in radioactive waste. In this study, we focused on the determination of 129I in cement waste form containing 137Cs, 60Co. We added 1 g of 129I(11.084 Bg), 137Cs(1,300 Bq) and 60Co(402 Bq) to cement waste form, respectively. The separation of 129I in cement waste form was carried out using an acid leaching method. And, we confirmed the specific activity of 137Cs and 60Co at each separation step. It was observed that an acid leaching method showed the remove efficiency 137Cs(99.97%) and 60Co(99.94%), respectively. In addition, 129I was also analyzed at approximately 96.44% in simulated contaminated cement waste form. In conclusion, through this experiment, it was confirmed that 129I could be successfully separated and analyzed by using the acid leaching method in cement waste form containing excessive 137Cs and 60Co.

Mobility of radionuclides (RNs) in natural water systems can be increased by complex formation with organic materials. In alkaline cement pore-water conditions, cellulose materials in radwastes such as woods and papers are degraded fast to small organic materials. As a major cellulose degradation product, isosaccharinate (ISA) has been paid attention recently due to its effect on facilitating RNs migration. ISA contains a carboxyl and four hydroxyl functional groups, which cooperatively interact to form chelating bonds with positively charged radionuclides. In our previous study, we determined thermodynamic formation constants, reaction enthalpy and entropy of trivalent americium complexes with ISA, Am(ISA)n (3-n)+ (n=1, 2), in weak acidic condition by conducting temperature-dependent UVVis absorption spectroscopy. Based on those thermodynamic constants along with the experimental results from time-resolved laser induced fluorescence spectroscopy and DFT calculations, we suggested two different chelating-modes of ISA on Am(III). It is more relevant to study Am(III)-ISA complexation under alkaline conditions around pH 12.5, which correspond to the pore-water condition of calciumsilicate- hydrate. Under the alkaline conditions, deprotonated hydroxyl groups of ISA can form more strong interactions with Am. Aquatic hydroxide group can also act as a ligand to form ternary Am(III) -ISA-OH complexes. In this study, absorption spectra of Am-ISA systems were monitored with two variations: first, pH variation (5.5–13) in the presence of constant 30 mM ISA, and second, ISA concentration variation (20 μM – 30 mM) at constant pH of 12.5. As increasing the pH at constant 30 mM ISA, absorption spectra of Am(ISA)2 + were red-shifted from 506.3 to 509.5 nm. The samples showed stable absorption spectra over 30 days. On the other hand, samples with lower ISA concentrations below 10 mM at pH 12.5, showed gradual decrease in the absorbance as sample aging time. By examining filtrates after ultrafiltration (1 kDa), we confirmed that aqueous Am(III)-ISA complexes were formed in the presence of 30 mM ISA at pH 12.5, while colloidal particles and precipitations were formed in the conditions of ISA concentrations lower than 10 mM. In this presentation, we will discuss about probable ternary complex forms of Am(III)-ISA-OH, colloidal forms, and solubility of Am(III) as a function of ISA concentration under alkaline conditions. Absorption and luminescence spectroscopic properties of the Am(III)-ISA-OH ternary system will also be presented.

Solubility and species distributions of radionuclides in domestic groundwater conditions are required for the safety assessment of deep underground disposal system of spent nuclear fuel (SNF). Minor actinides including Am contribute significant extents to the long-term radiotoxicity of SNF. In this study, the solubility of Am was evaluated in synthetic groundwater (Syn-DB3), which were simulated for the groundwater of the DB3 site in the KAERI Underground Research Tunnel (KURT). Geochemical modeling was performed based on the ThermoChimie_11a (2022) thermochemical database from Andra to estimate the solubility and species distributions of Am in the Syn-DB3 condition. Dissolved Am concentrations in the Syn-DB3 were experimentally measured under oversaturation conditions. Am(III) stock solution in perchlorate media was sequentially diluted in Syn-DB3 to prepare 8 μM Am(III) in Syn-DB3. The pH of the solutions was adjusted to be in the range of 6.4–10.5. A portion of the samples was transferred to quartz cells for UV-Vis absorption and time-resolved laser fluorescence spectroscopy studies and the rest were stored in centrifuge tubes. The absorption spectra of the samples were monitored over 70 days and the results suggest that Am colloidal particles were formed initially in all the samples and precipitated rapidly within two days. Over the experimental period of 236 days, small volume (10 μL) of the samples in the centrifuge tubes were periodically withdrawn after centrifugation (18000 rpm, 1 hr) for the liquid scintillation counting to measure the concentrations of Am dissolved in Syn-DB3. In the end of the experiments, pH of the samples was checked again and the final dissolved Am concentrations were determined after ultrafiltration (10 kDa) to exclude the contribution of colloidal particles. In the pH range of 8-9, which is relevant to the KURT-DB3 groundwater condition, the measured dissolved Am(III) concentrations were converged to around 10-8 M. These values are higher than the solubility of AmCO3OH:0.5H2O(s), but lower than that of AmCO3OH(am). There was no indication of transformation of the amorphous phase to the crystalline phase in our observation time window.

Zinc injection into the coolant system of nuclear power plants is an effective method for reducing corrosion and improving performance. The effectiveness of this method is influenced by various factors such as zinc concentration and injection rate. This paper provides an overview of the factors affecting the effectiveness of zinc injection in nuclear power plants, with a focus on zinc concentration and injection rate, and discusses various research results on the effects of these factors on corrosion reduction and coolant system performance. Zinc concentration is an important factor affecting the effectiveness of zinc injection. The research results show that gradual increases in zinc concentration are more effective for coolant system stability. However, the concentration should not exceed the recommended levels as high zinc concentrations can have negative effects on the system. Injection rate is also an important factor affecting the effectiveness of this method. The research results show that gradual increases in injection rate are more effective for coolant system stability. However, excessive injection rates can have negative effects on the system such as overload of the zinc injection facility and chemical shocks within the coolant system, and therefore, should be optimized. In conclusion, zinc concentration and injection rate are important factors affecting the effectiveness of zinc injection in nuclear power plants. The optimal concentration and injection rate should be determined based on specific reactor conditions and system requirements, and efforts should be made to maximize corrosion reduction and performance improvement.

A low- and intermediate-level radioactive waste repository contains different types of radionuclides and organic complexing agents. Their chemical interaction in the repository can result in the formation of radionuclide-ligand complexes, leading to their high transport behaviors in the engineered and natural rock barriers. Furthermore, the release of radionuclides from the repository can pose a significant risk to both human health and the environment. This study explores the impact of different experimental conditions on the transport behaviors of 99Tc, 137Cs, and 238U through three types of barrier samples: concrete, sedimentary rock, and granite. To assess the transport behavior of the samples, the geochemical characteristics were determined using X-ray diffraction (XRD), X-ray fluorescence (XRF), Fouriertransform infrared spectroscopy (FTIR), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS), and Brunauer-Emmett-Teller (BET) analysis. The adsorption distribution coefficient (Kd) was used as an indicator of transport behavior, and it was determined in batch systems under different conditions such as solution pH (ranging from 7 to 13), temperature (ranging from 10 to 40°C), and with the presence of organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA). A support vector machine (SVM) was used to develop a prediction model for the Kd values. It was found that, regardless of the experimental parameters, 99Tc may migrate easily due to its anionic property. Conversely, 137Cs showed low transport behaviors under all tested conditions. The transport behaviors of 238U were impacted by the order of EDTA > NTA> ISA, particularly with the concrete sample. The SVM models can also be used to predict the Kd values of the radionuclides in the event of an accidental release from the repository.

Zirconium(Zr) alloys are commonly used in the nuclear industry for applications such as fuel cladding and pressure tubes. To minimize the levels and volumes of radioactive waste, molten salts have been employed for decontaminating Zr alloys. Recently, a two-step Zr metal recovery process, combining electrolysis and thermal decomposition, has been proposed. In the electrolysis process, potentiostatic electrorefining is utilized to control the chemical form of electrodeposits(ZrCl). Although Zr metals are expected to dissolve into molten salts, reductive alloy elements can also be co-dissolved and deposited on the cathode. Therefore, a better understanding of the anodic side’s response during potentiostatic electrorefining is necessary to ensure the purity of recovered Zr and long-term process operation. As the first step, potentiodynamic polarization curves were obtained using Zr, Nb, and Zr-Nb alloy to investigate the anodic dissolution behavior in the molten salts. Nb, which has a redox potential close to Zr, and Zr exhibit active or passivation dissolution mechanisms depending on the potential range. It was confirmed that Zr-Nb alloy also has a passivation region between -0.223 to -0.092 V influenced by the major elements Zr and Nb. Secondly, active dissolution of Zr-Nb was performed in the range of -0.9 to -0.6 V. The dissolution mechanism can be explained by percolation theory, which is consistent with the observed microstructure of the alloy. Thirdly, passivation dissolution of Zr, Nb, and Zr-Nb alloy was investigated to identify the pure passivation products and additional products in the Zr-Nb alloy case. K2ZrCl6 and K3NbCl6 were identified as the pure passivation products of the major elements. In the Zr-Nb alloy case, additional products, such as Nb and NbZr, produced by the redox reaction of nanoparticles in the high viscous salt layer near the anode, were also confirmed. The anodic dissolution mechanism of Zr-Nb alloy can be summarized as follows. During active dissolution, only Zr metal dissolves into molten salts by percolation. Above the solubility near the anode, passivation products begin to form. The anode potential increases due to the disturbance of passivation products on ion flow, leading to co-dissolution of Nb. When the concentration of Nb ion exceeds the solubility, a passivation product of Nb also forms. In this scenario, a high viscous salt layer is formed, which traps nanoparticles of Zr metal, resulting in redox behavior between Zr metal and Nb ion. Some nanoparticles of Zr and Nb metal are also present in the form of NbZr.

Support for nuclear power plant (NPP) dismantling & decommissioing (D&D) industry is necessary through development of the infrastructure and the D&D technology. Because KORI#1 and Wolsong#1 is planned to decommission until around 2030. Korea research institute of decommissioing (KRID) was established through the preliminary feasibility study. KRID has plan to support nuclear companies to join D&D industry. Normal facilities (Lv.1) of KRID infracstucture are currently being constructed and radiation management facilities (Lv.2) construction is expected to begin in October. Further, KRID is planning the construction of equipment to develop the procedure for radionuclide analysis through R&D project. A total period of the R&D project is 45 months, and the total R&D funding for this period is 19.4 billion won. The ultimate goal of the R&D project is to build the infractstucture base to analyze decommissioning radioactive wastes. Furthermore, the R&D project is important to reliably perform the NPP D&D.

Molten salt reactors have several advantages over conventional light water reactors. These include producing less nuclear waste, operating at higher power efficiency and inherent safety due to the low operating pressure. NaCl-MgCl2 eutectic salt is one of the candidates for the molten salt reactor coolant. However, because the salt is very hygroscopic, structural material corrosion occurs resulting in the high cost to maintain. To mitigate corrosion there have been many studies for the dehydration of the salt, especially focusing on the magnesium chloride. The reason is that the moisture adsorbed to the magnesium chloride undergoes hydrolysis over 200 degrees Celsius and decomposes to MgOHCl while the moisture associated with the NaCl is easily liberated during the heating procedure without chemical reaction. As the operating temperature of the molten salt is between 500 and 700 degrees Celsius, the MgOHCl is believed as the main cause for the structural corrosion. In this research, thermal dehydration of the salt with elemental Mg, for the NaCl-MgCl2 eutectic, was studied based on the previous dehydration methods and considering scalable and easy to handle. The MgOHCl was removed both through the thermal decomposition and the reduction by Mg metal. After the removal of MgOHCl, based on the difference between the freezing points and the density, the salt cooled down very slowly to ensure the separation between the purified salt and the disposals such as MgO and remaining Mg metals. The efficiency of the dehydration method was determined by the concentration of the MgOHCl. The concentration was determined by cyclic voltammetry and the result was compared with undehydrated salt and salt dehydrated thermally without the addition of Mg metal. To qualify and quantify the MgOHCl content through the cyclic voltammetry, it was necessary to observe the signal by adding MgOHCl to each sample. Based on the thermogravimetric analysis result of MgCl2· 6H2O, MgOHCl powder was formed through heating the MgCl2·6H2O.

Change in chemical constitution of a zirconium sample treated at the simulated condition of a pressurized water reactor (water at 315°C and 15.5 MPa) was investigated using X-ray diffraction (XRD) and Raman spectroscopy. We observed swelling of the zirconium sample as well as the change in its color from silver to gray treated after the pressurized water at high temperature. On the basis of XRD and Raman data, we confirmed that the variation in composition of zirconium specimen from hexagonal Zr to monoclinic ZrO2 occurred at the simulated PWR condition. Therefore, we suggested that the oxidation of zirconium appeared due to the reaction with water at high temperature and pressure, as shown in the following reaction: Zr + 2H2O → ZrO2 + 2H2.

Confirmation of the thermal behavior of spent fuel is one of the important points in the management of high-level radioactive waste. This is because various fission products exist in spent nuclear fuel, and a management plan according to their behavior is required. Among the fission products, epsilon particles exist in the form of metal deposits and have a great influence on their physical and chemical properties. However, observing the thermal behavior of epsilon particles is important for understanding spent fuel behavior in thermally environment, but it is difficult to maintain a consistent thermal environment. In this work, we report the thermal behaviors study of uranium oxide with epsilon particle using in situ high temperature X-ray diffraction. We measured the variation of temperature on the size of crystalline, which is a cell parameter in the reaction process. And then, the change of lattice parameters is calculated by Rietveld refinement.

Viscosity of molten salts is an essential property for the thermal hydraulic design and evaluation of molten salt reactor (MSR). Therefore, viscosity data is one of the fundamental physical property data required for safe process operation and countermeasures to severe accidents. In this study, based on our experience of developing a viscosity measurement system for high-temperature LiCl-KCl molten salt system, the viscosity of NaCl-MgCl2 and NaCl-MgCl2-UCl3 molten salts, which are considered promising salts in MSR, was measured. In order to investigate the physical properties of uranium in high-temperature NaCl-MgCl2 molten salt, a viscometer system for high-temperature viscosity measurement was specially designed. As a result of the measurement, the viscosity of the 58NaCl- 42MgCl2 molten salt was 2.73 cP at 838 K, 2.15 cP at 889 K, and 1.68 cP at 940 K. And the viscosity of 73NaCl-21MgCl2-6UCl3 molten salt was 3.79 cP at 877 K, 3.58 cP at 897 K, and 1.63 cP at 941 K. The repeatability of the measurement showed a precision of less than 3%. Although sufficientlyverified starting materials were not used, viscosity data were reported for the first time for NaCl- MgCl2-UCl3 molten salts.

In molten salt reactor (MSR), liquid fuel integrated with the molten salt coolant is used to improve the safety, resulting in the prevention of the loss-of-coolant accident (LOCA) that can occur in a pressurized water reactor (PWR). Because the structural materials used in MSRs are directly contacted with liquid fuel for a long time, they must have excellent corrosion resistance to the molten salt. Therefore, to examine the corrosion rates for Ni-based alloys in the molten salt, the corrosion experiments for alloy 600, alloy 617, and Hastelloy N were performed under LiCl molten salt at 635°C for 100 h in a glove box under Ar environment. Through a weight loss method for the three Nibased coupons before and after the corrosion tests, we evaluated their corrosion rates. Based on the results of weight loss for each alloy, we confirmed that Hastelloy N has the excellent corrosion resistance compared to the other alloys. Furthermore, the changes in the crystal structure and surface morphology with elemental distribution for the three alloys by corrosion in LiCl molten salt were analyzed, showing the variation in surface topography and the decrease in Cr element after corrosion experiments for all coupons.

The density of molten salts is the most important property in the development of molten salt reactor (MSR). The density value measured through the experiment is also very valuable as a gold standard for the validation of the prediction models based on molecular dynamics or other computational methods. To the best of our knowledge, the experimental density data of the ternary NaCl-MgCl2- UCl3 salt system as a MSR candidate fuel salt have never been reported previously. In this study, density measurement experiment of high-temperature molten salt of NaCl-MgCl2 and NaCl-MgCl2- UCl3 was conducted using a previously-developed density measurement system based on the maximum bubble pressure (MPB) method. As a result of the experiment, the density value of 62NaCl- 18MgCl2-20UCl3 molten salt at 873 K was 2.62 g/cm3. A density prediction value of 2.65 g/cm3 at 873 K was derived from the obtained results based on the rule of additivity of molar volume method. The predictred density of 62NaCl-18MgCl2-20UCl3 was consistent with the experimental value within 1%. The density measuring system used in this study is promising for the validation of other multicomponent molten salt systems.

Molten salt is one of the promising medium materials for molten salt reactors and energy storage systems. Molten salt is advantageous for better physical properties such as low melting point and high boiling point, high energy capacity, high thermal conductivity, and high thermal stability than other medium materials such as water or liquid metals. However, the corrosivity of the molten salt is one of the main factors that disturbs the various applications of the molten salt. On the other hand, metallic 3-D printing technologies have developed by leaps and bounds over the past 20 years and show potential for use in cutting-edge industries such as aerospace and military purposes. However, the biggest problem of 3-D printed products is that the mechanical and physical properties are very weak along the laminated plane that was generated during the manufacturing process. In particular, other research showed that corrosion is vulnerable through the laminated surface, and corrosion along the laminated plane is not completely mitigated through a general heat treatment process although the microstructure of the surface is evaluated to be partially mitigated by the heat treatment. In this study, molten salt corrosion behaviors of simple Ni-based alloy with a composition of 80Ni- 20Cr were analyzed. Ni-based alloys were fabricated by casting and 3-D printing, and some of the 3-D printed specimens were thermally treated at 1,273 K for 1 hour to examine the effects of heat treatment on corrosion behaviors. In molten eutectic NaCl-MgCl2 melts at 973 K, Ni-based alloys were corroded for 1, 3, 7, and 28 days and their microstructural changes were analyzed by SEM-EBSD-EDS and OM. The corrosion behaviors of the alloy were also evaluated by the salt composition measured with ICPOES. 3-D printed alloy with post-treatment showed more resistivity to the molten salt corrosion than as-fabricated 3-D printed alloy. However, the corrosion rate of the 3-D printed specimen after heat treatment was still higher than that made by casting.

Plutonium exhibits a variety of oxidation states and has a strong affinity for complexation with organic ligands. Isosaccharinic acid (ISA) is a major degradation product of cellulose materials present in the low to intermediate radioactive wastes. The interaction between trivalent plutonium and ISA can significantly impact the migration and containment of plutonium in the repository environment. In this study, formation of Pu(III) and ISA complexes was investigated at an ionic strength of 1 M of NaClO4 using UV-Vis absorption spectrophotometry. To exclude the effect of the Pu(III) oxidation, absorption spectra were measured within 10 min after adding ISA into Pu(III) solution and processed using HYPSPEC software for deconvolution after baseline correction. Several previous studies showed that the presence of ligands accelerates the oxidation of Pu(III) to Pu(IV). To investigate whether ISA complexation can also accelerate the Pu(III) oxidation, UV-Vis absorption spectra changes over 24 hours were analyzed as a function of the ratio of ISA to plutonium concentration.

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.