Wasteform is the first barrier to prevent radionuclide release from repositories into the biosphere. Since leaching rates of nuclides in wasteform significantly impact on safety assessment of the repository, clarifying the leaching behavior is critical for accurate safety assessment. However, the current waste acceptance criteria (WAC) of the Gyeongju repository only evaluates leachability indexes for Cs, Sr, and Co, which are tracers for nuclear power plant waste streams. Furthermore, ANS 16.1, the current leaching test method used in WAC, applies deionized water (DI) as leachant. However, the interactions between wasteform and groundwater environment in the repository may not be reflected. Therefore, it is necessary to review the current leaching test method and nuclides that may require the extra evaluation of leachability beyond the Cs, Sr, and Co. Tc and I are key nuclides contributing to high radioactive dose in safety assessment due to their high mobility and low retardation factor. The groundwater conditions within the repository, such as pH and Eh significantly affect the chemical form of Tc and I. For example, Tc in H2O system tends to form hydroxide precipitates in neutral pH condition and TcO4 - in strong alkaline environments according to the Pourbaix diagram. In case of I, it generally exists in the form of I-, while it exists as IO3 - as Eh increases. Although the current leaching test at the Gyeongju repository applies DI as a leachant, the actual repository is expected to have a highly alkaline environment with a substantial amount of various ions in the groundwater. Consequently, the leaching behavior in the ANS 16.1 test and the actual disposal condition is different. Thus, it is necessary to analyze the leaching behavior of Tc and I with reflecting the actual disposal environment. In this study, the leaching behavior of Tc and I is investigated by following ANS 16.1 leaching test method. The solidified waste specimens containing 10 mmol of Re and I were manufactured with cement, which is widely used as a solidification material. Re was applied instead of Tc, which has similar chemical behavior to Tc, and NH4ReO4 and NaI were used as surrogates for Re and I. As a leachant, deionized water and cement-saturated groundwater were prepared and the concentration of nuclides in the leachant is analyzed by ICP-OES. As the result of this study, experimental data can be applied to improve the WAC and disposal concentration standards in the future.

Radionuclides in low- and intermediate-level radioactive wastes from the decommissioning process of nuclear power plants were generally immobilized by cementation methods. Ethylenediaminetetraacetic acid (EDTA), which is extensively used as a decontamination agent, can affect the behaviors of radionuclides immobilized in cement waste forms. In this study, the effects of EDTA contained in simulated radioactive decommissioning wastes on the leaching characteristics of immobilized Co and Cs and the microstructure evolution of cement waste form. Co leaching was accelerated by the formation of Co–EDTA complexes with high mobility and solubility. Cs leaching was hindered by the ion competition with other metal–EDTA complexes for releasing from the cement waste form. Cs leaching was also retarded by carbonated layer at edge of the cement waste form, which process is facilitated by the presence of EDTA. Finally, the effects of EDTA on the leaching characteristics of immobilized Cs and Co and the microstructure evolution of the cement waste form should be considered to ensure the safety of disposal for lowand intermediate-level radioactive wastes.

In this study, we evaluated the effects of acid leaching on the properties of Cr powder synthesized using self-propagating high-temperature synthesis (SHS). Cr powder was synthesized from a mixture of Cr2O3 and magnesium (Mg) powders using the SHS Process, and the byproducts after the reaction were removed using acid leaching. The properties of the recovered Cr powder were analyzed via X-ray diffraction (XRD), scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), particle size analysis (PSA), and oxygen content analysis. The results show that perfect selective leaching of Cr is challenging because of various factors such as incomplete reaction, reaction kinetics, the presence of impurities, and incompatibility between the acid and metal mixture. Therefore, this study provides essential information on the properties under acidic conditions during the production of high-quality Cr powder using a self-propagating high-temperature synthesis method.

Several tests should be performed to estimate the structural and chemical stability of the radioactive waste. Among the tests in Gyeongju LILW repository, the leaching test which follows ANS 16.1 standard test method should be conducted for Cs, Sr, and Co radionuclides and must satisfy leachability index larger than 6 which applies deionized water as a leachant. However, the expected leachant inside the silo is groundwater that contains various ions and a high pH condition is predicted due to the concrete structures inside the silo. According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. Cobalt precipitates when the leachant has high pH environment which is similar condition to the cement-saturated leachant. Unlike the cobalt, cesium is preferred to exist as ion in the high pH condition. Therefore, the significant effect of the chemical environment of the leachant on the leachability of the radionuclides should be considered for the waste acceptance criteria of the radioactive waste repository. From the ‘NRC, Technical position on the waste form, rev1’, the leaching test method should follow the ANS 16.1 methods by using deionized water as leachant, however, a new leachant showing more aggressive leachability can be applied instead of deionized water. In the other hand, ASTM C1308 leaching test method recommends applying actual groundwater of the repository as a leachant. FT-04-020, the leaching test method of France, suggests the ion composition of the groundwater including the pH value. Therefore, the adequacy of using deionized water as leachant for the leaching test method of Cs, Sr, and Co should be re-examined. In this study, the leaching behavior of Cs, Sr, and Co under the several leachant types is estimated. The cement solidified specimen containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water, simulated groundater, and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type was discussed. The result of this study is expected to be a background data that helps understanding the actual leaching behavior of the Cs, Sr, and Co in the Gyeongju LILW repository.

Many countries have used nuclear power to generate electricity. Uranium-235, which is used as fuel in nuclear power plants, produces many fission products. Among them, iodine-129 is problematic due to its long half-life (1.57×107 years) and high diffusivity in the environment. If it is released into the environment without any treatment, it could have a major impact on humans and ecosystems. Therefore, it must be treated into a stable form through capture and solidification. Iodine can be captured in the form of AgI through silver-loaded zeolite filters in off-gas treatment processes. However, AgI could be decomposed in the reducing atmosphere of groundwater, so it must be converted into a stable form. In this study, Al2O3, Bi2O3, PbO, V2O5, MoO3, or WO3 were added to the iodine solidification matrix, AgI-Ag2O-TeO2 glass. The glass precursors were mixed to the appropriate composition and placed in an alumina crucible. After heat treatment at 800°C for 1 hour, the melt was quenched in a carbon crucible. The leaching behavior and thermal properties of the glass samples were evaluated. The PCT-A test for leaching evaluation showed that the normalized releases of all elements were below 2 g/m2, which satisfied the U.S. glass wasteform leaching regulations. Diffrential scanning calorimetry (DSC) was performed to evaluate the thermal properties of all glass samples. The addition of MoO3 or WO3 to the AgI-Ag2O-TeO2 glass increased the glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability. The similar relative electro-static filed values of MoO3, and WO3 which are approxibately three times that of the glass network former TeO2, could provide sufficient force to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M=V, Mo, W) links. The high electrostatic forces of Mo and W increased the glass network cohension and prevented the crystallization of the glass.

Heat-generating nuclides such as Cs-137 and Sr-90 should be separated from spent nuclear fuel to reduce the short-term thermal load on the repository facility. In particular, Sr-90 must be separated because its decay process generates high temperatures. Recently, the Korea Atomic Energy Research Institute (KEARI) has been developing a waste burden minimization technology to reduce the environmental burden resulting from the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology incorporates a nuclide management process that maximizes disposal efficiency by selectively separating and accumulating key nuclides from spent nuclear fuel, such as Cs, Sr, I, TRU/RE, and Tc/Se. Sr nuclides dissolve in the chloride phase during the chlorination process of spent nuclear fuel and are recovered as carbonate or oxide through reactive distillation or reactive crystallization. Due to their chemical similarity, Ba nuclides are recovered along with Sr nuclides during this process. In this study, we prepared a ceramic waste form for group II nuclides, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method, taking into account the different ratios of Sr/Ba nuclides produced during the nuclide management process. Regardless of the Sr/Ba ratio, the established waste form fabrication process was able to produce a stable waste form. Physicochemical properties, including leaching and thermal properties, were evaluated to determine the stability of group II waste forms. In addition, the radiological properties of waste forms of Ba(x)Sr(1-x)TiO3 with varying Sr/Ba ratios were evaluated. These results provided fundamental data for the long-term storage and management of waste forms containing group II nuclides.

The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

Irradiated uranium dioxide in damaged used fuel could oxidize during transportation, interim storage or disposal, resulting that the fuel pellet fragments are reduced to a grain-sized powder that can easily escaped from the damaged rod. It has been reported that oxidized spent fuel (i.e. U4O9+x) that was in contact with water could increase the dissolution rate by making the grain boundaries more accessible to the water. Therefore, the damaged used fuel requires stabilization technology including nuclear material recovery, pellet manufacturing process, and stabilization fuel rod manufacturing that can secure safety in terms of permanent disposal. In this study, we prepared pure UO2 and SIMFUEL pellets that are a mixture of UO2 and surrogated metallic oxides for fission products equivalent to a burn-up of 35 GWd/tU and 55 GWd/tU as the stabilized spent fuel. The UO2 and fission products powders were milled and pressed into pellets at 250 MPa and sintered at 1,550°C and 1,700°C for 6 hours in an atmosphere of 4%H2-Ar. The prepared UO2 and SIMFUEL pellets were placed in PTFE Teflon vessels and filled with deionized water to identify the leaching behavior by a long-term leaching experiment under the similar condition to a repository for the safe disposal.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

Numerous low-and intermediate level radioactive wastes were generated from the decommissioning processes of nuclear power plants. Radionuclides such as Co and Cs contained in decommissioning wastes should be immobilized to prevent the release of radionuclides from the wastes due to its harmful impacts on ecosystem by high radioactivity and long half-life. Ethylenediaminetetraacetic acid (EDTA) used as decontamination agent can be contained in cement waste during decommissioning process of nuclear power plants. In addition, EDTA can be stably and strongly bound with radionuclides, resulting in the acceleration of the nuclide release from solidified cement matrix. Here, we investigated the effects of EDTA on leaching behaviors of Co and Cs immobilized in the cement specimen. The leaching tests were performed according to the ANS 16.1 “Measurement of the leachability of solidified low-level radioactive wastes by a short-term test procedure”. From the results, an increase in the EDTA content in the cement specimen led to an increase in Co leaching, whereas a decrease in Cs leaching. Leaching of Cs was dominantly controlled by diffusion from the pore space of the cement specimen to the solution. The effective diffusion coefficient and leachability index of nuclide were determined using the diffusion-release models of ANS 16.1. The results of present study can be used in the safety assessment for disposal of the radioactive waste generated by decommissioning of nuclear power plants.

Garnet is one of the promising ceramic waste forms for immobilizing radioactive wastes. It has an A3 [VIII]B2 [VI]T3 [IV]O12 structure, so it can accommodate various cations of different sizes and coordination. Silicon usually occupies the centers of the tetrahedron structural site (T[IV]O4) in natural garnet. However, substitution of the T-site with iron, which has a relatively large ionic radius, causes the expansion of a unit cell volume of garnet and allows the incorporation of large cations such as actinides at other sites. Relatively few leaching data have been reported for ferrite garnet waste forms to date. In this study, we synthesized gadolinium-iron-garnet and evaluated the leaching property using cerium as a surrogate for actinide elements. The test specimens were made by cold pressing and sintering process. Three different standard leaching tests were performed as follows. The PCT-A (ASTM C1285) was performed for 7 days at 90°C to the crushed sample (0.149 to 0.074 mm). The ANSI/ANS-16.1 standard leach test was performed at ambient conditions for 5 days with constant replacement of leachate. Finally, the MCC-1 (ASTM C1220) test was performed for 28 days at 90°C with different types of leachants such as ultrapure water, brine, and silicate water. The last two leaching tests were conducted on monolithic specimens. After the end of the test, leachate was analyzed by inductively coupled plasma mass spectroscopy (Agilent, ICP-MS 7700S).

Tungsten carbide is widely used in carbide tools. However, its production process generates a significant number of end-of-life products and by-products. Therefore, it is necessary to develop efficient recycling methods and investigate the remanufacturing of tungsten carbide using recycled materials. Herein, we have recovered 99.9% of the tungsten in cemented carbide hard scrap as tungsten oxide via an alkali leaching process. Subsequently, using the recovered tungsten oxide as a starting material, tungsten carbide has been produced by employing a self-propagating high-temperature synthesis (SHS) method. SHS is advantageous as it reduces the reaction time and is energy-efficient. Tungsten carbide with a carbon content of 6.18 wt % and a particle size of 116 nm has been successfully synthesized by optimizing the SHS process parameters, pulverization, and mixing. In this study, a series of processes for the highefficiency recycling and quality improvement of tungsten-based materials have been developed.

Cobalt (Co) is mainly used to prepare cathode materials for lithium-ion batteries (LIBs) and binder metals for WC-Co hard metals. Developing an effective method for recovering Co from WC-Co waste sludge is of immense significance. In this study, Co is extracted from waste cemented carbide soft scrap via mechanochemical milling. The leaching ratio of Co reaches approximately 93%, and the leached solution, from which impurities except nickel are removed by pH titration, exhibits a purity of approximately 97%. The titrated aqueous Co salts are precipitated using oxalic acid and hydroxide precipitation, and the effects of the precipitating agent (oxalic acid and hydroxide) on the cobalt microstructure are investigated. It is confirmed that the type of Co compound and the crystal growth direction change according to the precipitation method, both of which affect the microstructure of the cobalt powders. This novel mechanochemical process is of significant importance for the recovery of Co from waste WC-Co hard metal. The recycled Co can be applied as a cemented carbide binder or a cathode material for lithium secondary batteries.

An important property of glass and ceramic solid waste forms is processability. Tellurite materials with low melting temperatures and high halite solubilities have potential as solid waste forms. Crystalline TiTe3O8 was synthesized through a solid-state reaction between stoichiometric amounts of TiO2 and TeO2 powder. The resultant TiTe3O8 crystal had a three-dimensional (3D) structure consisting of TiO6 octahedra and asymmetric TeO4 seesaw moiety groups. The melting temperature of the TiTe3O8 powder was 820℃, and the constituent TeO2 began to evaporate selectively from TiTe3O8 above around 840℃. The leaching rate, as determined using the modified American Society of Testing and Materials static leach test method, of Ti in the TiTe3O8 crystal was less than the order of 10-4 g·m-2·d-1 at 90℃ for durations of 14 d over a pH range of 2-12. The chemical durability of the TiTe3O8 crystal, even under highly acidic and alkaline conditions, was comparable to that of other well-known Ti-based solid waste forms.