Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

아메리카동애등에 성충은 음식물 폐자원 등 유기물이 있는 곳에 알을 낳는 습성이 있다. 대부분의 농가는 음식 물폐자원을 가공한 단미사료(습식사료)를 유인배지로 활용하여 그 위에 플로랄폼(오아시스)를 놓고 알을 받는 다. 그러나 플로랄폼은 재사용이 불가하고 생분해되지 않는 환경폐기물로서 처리가 곤란하며 포름알데하이드, 카본 블랙 등의 발암물질을 함유한 것으로도 알려져 있다. 이에 본 연구는 먹이원 자체를 활용하여 폐기물이 발생하지 않는 친환경 산란받이를 개발하였으며 일회용으로 사용되는 플로랄폼을 대체하였다. 먹이원으로 활 용할 수 있는 습식사료와 건식사료를 주재료로 하여 제작하며, 습식사료(수분60~80%)와 건식사료(1~10%)를 1:0.5~1 비율로 혼합한 사료 혼합물과 보조첨가제와 물을 포함하여 제작한다. 친환경 산란받이는 기존 플로랄폼 대비 산란율이 34% 증가하였으며 구매비용 또한 75% 절감하였다.

담배거세미나방(Spodoptera litura), 열대거세미나방(Spodoptera frugiperda) 및 파밤나방(Spodoptera exigua) 은 광식성 해충이지만, 종특이적인 기주범위를 갖는다. 이들이 기주식물을 찾아가는 과정에 냄새감각이 어떤 역할을 하는지 알아보기 위해, 기주 및 비기주 식물에서 발산되는 46가지 휘발성 물질을 선정하여, 이들에 대한 나방 3종의 냄새반응을 GC-EAD(gas chromatography-electroantennogram detection)를 통해 확인하였다. 그 결과, 46가지의 식물유래화합물 중 9가지 물질이 3종의 나방 모두에서 냄새활성을 나타냈으며, 2가지 물질은 담배거세 미나방과 열대거세미나방에만 냄새활성을 나타냈고, 다른 몇 가지 물질은 담배거세미나방에만 냄새활성을 나 타냈다. 이 결과는 세 종 나방이 식물냄새물질 탐지를 위해 유사한 냄새감각세포를 가지며, 일부 종에서는 종특이 적인 냄새감각세포가 존재한다는 것을 보여준다. 이 결과를 바탕으로 냄새활성을 나타낸 물질들의 나방 3종에 대한 행동활성을 야외트랩실험을 통해 확인할 예정이다.

Sweet pepper(paprika) belongs to the genus Capsicum, and is one of the most important export product from Korea to Japan and Southeast Asia. So it is important to eradicate plant quarantine pests before export sweet pepper. Aphids, whiteflies and mites are major pests that can damage to sweet peppers. Fumigation is normally used to eradicate pests in plant quarantine, but phytotoxicity may can be appeared that affect the quality of the product. Low-temperature treatment, one of the most popular physical treatment, can reduce crop damage to preserve product quality, but it takes long time to kill pests, which can cause quality degradation. In this study, phytotoxicity of fumigants, phosphine(PH3), ethyl formate(EF) and PH3+EF on sweet peppers was investigated to use as basic data for physicochemical treatment. When treated with more than 35 mg/L of EF, phytotoxicity was occurred, and was not occurred with PH3. When low-temperature of 1.7 degrees treated for 15 days after fumigation, it seems to be no direct damage from low-temperature treatment. But quality of top of sweet pepper was decreased from 7 days after fumigation.

Mass production of high-quality carbon nanotubes (CNTs) remains a challenge, requiring the development of new wetimpregnated catalyst suitable for the catalytic chemical vapor deposition (CCVD) of CNTs in a fluidized bed reactor. For the successful development of a new catalyst, a highly robust system to synthesize CNTs must be established. Here, we systematically investigated the robustness of CNT synthesis by CCVD using a wet-impregnated catalyst. We statistically tested four factors that could potentially affect the robustness of CNT synthesis system, focusing on carbon yield and IG/ID. First, we tested the effect of vacuum baking before CNT growth. F test and CV equality test concluded that vacuum baking recipe did not significantly reduce the variability of the CNT synthesis. Second, we tested the batch-to-batch variation of catalysts. The results of t test and one-way analysis of variance indicate that there is significant difference in carbon yield and IG/ID among catalysts from different batches. Third, we confirmed that there is spatial non-uniformity of wet-impregnated catalysts within a batch when they are produced in large scale. Finally, we developed a multi-step heating recipe to mitigate the temperature overshooting during the CNT synthesis. The multi-step recipe increased the mean of carbon yield, but did not influence the variability of CNT synthesis. We believe that our research can contribute to the establishment of a robust CNT synthesis system and development of new wet-impregnated catalysts.

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.

The separation of hydrogen isotopes is a critical issue in various fields, such as deuterium or tritium production and the treatment of radioactively contaminated water. In this presentation, we describe the pervaporative separation of hydrogen isotopes using proton conductive membranes and underlying separation mechanism. We investigated the H/D separation factors of perfluorosulfonic acid (Nafion) and polybenzimidazole membranes using pervaporation, and found that both membranes exhibited similar separation factors of approximately 1.026. Water permeation flux through the membranes was highly dependent on their thickness and type, and increased with operation temperature. However, the effect of temperature on H/D separation factor was negligible. We also demonstrated the cascade separation of H/D, indicating the potential application of multi-stage operation. We found that surface transport mechanisms such as hydron hopping contributed the most to H/D separation during the pervaporation process of proton conductive membranes.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

According to IAEA PRIS, there is no record of dismantling commercial heavy water reactors among 57 heavy water reactors around the world. In Canada, which has the largest number of heavy water reactors, three of the 22 commercial heavy water reactors with more than 500 MW are permanently suspended, Gentilly unit 2 (2012), Pickering unit 2 (2007), and Pickering unit 3 (2008), all of which chose a delayed decommissioning strategy. On the other hand, Wolsong unit 1, which will be the world’s first heavy water reactor to be dismantled commercially, will be immediately carried out as a decommissioning strategy. KHNP has established various cooperation systems with advanced companies and international organizations related to overseas NPP decommission and is actively exchanging technologies. Among them, the most important focus is on research cooperation related to COG (CANDU owners Group). The first case is a joint study on Conceptual Calandria Segmentation. Four areas of process, waste management, ALARA, and cost for decommissioning reactors to be submitted to Canadian regulators for approval of Pickering and Gentilly-2’s preliminary decommissioning plan have been evaluated, and research on Wolsong unit 1 is currently underway. The second case is Decommissioning and long-term waste management R&D. Although the technical maturity is low, it studies the common interests of member companies in the decommissioning of heavy water reactor power generation companies and long-term waste management. Robotics for dismantling high-radiation structures, C- 14, H-3 measurement and removal methods, and concrete decontamination technology, which are characterized by heavy water, are being actively studied. KHNP is strengthening international cooperation with COG to prepare for the successful decommissioning of Wolsong unit 1. Based on previous studies by Pickering and Gentilly-2, an evaluation of the decommissioning of Wolsong unit 1 reactor is being conducted. In addition, it is preparing for decommissioning through experience analysis of the pressure tube replacement project.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

The physicochemical similarities of hydrogen isotopes have made their separation a challenging task. Conventional methods such as cryogenic distillation, Girdler sulfide process, chromatography, and thermal cycling absorption have low separation factors and are energy-intensive. To overcome these limitations, research has focused on kinetic quantum sieving (KQS) and chemical affinity quantum sieving (CAQS) effects for selective separation of hydrogen isotopes. Porous materials such as metal-organic frameworks (MOF), covalent organic frameworks (COF), zeolites, carbon, and organic cages have been studied for hydrogen separation. This study have the literature review for previous research on D2/H2 adsorption and analyzes the D2/H2 adsorption behaviors of hydrogen isotopes for various zeolite using BET at 77 K. The study predicts the D2/H2 adsorption selectivity based on the results obtained with BET. These hydrogen isotope adsorption fundamentals provide a foundation for future processes for tritium separation.

With the recent concern regarding cellulose enhancing radionuclide mobility upon its degradation to ISA, disposal of cellulosic wastes is being held off until the disposal safety is vindicated. Thus, a rational assessment should be conducted, applying an appropriate cellulose degradation model considering the disposal environment and cellulose degradation mechanisms. In this paper cellulose degradation mechanisms and the disposal environment are studied to propose the best-suitable cellulose degradation model for the domestic 1st phase repository. For the cellulose to readily degrade, the pH should be greater than 12.5. As in the case of SKB, 1BLA is excluded from the safety assessment because the pH of 1BLA remains below 12.5. Furthermore, despite cellulose degradation occurring, it does not always produce ISA. At low Ca2+ concentration, the ISA yield rate is around 25%, but at high Ca2+ concentration, the ISA yield rate increases up to 90%. Thus, for the cellulose to be a major concern, both pH and Ca2+ concentration conditions must be satisfied. To satisfy both conditions, the cement hydration must be in 2nd phase, when the porewater pH remains around 12.5 and a significant amount of Ca2+ ion is leaching out from the cement. However, according to the safety evaluation and domestic research, 2nd phase of cement hydration for silo concrete would achieve a pH of around 12.4, dissatisfying cellulose degradation condition like in 1BLA. Thus, cellulose degradation would be unlikely to occur in the domestic 1st phase repository. To derive waste acceptance criteria, a quantitative evaluation should be conducted, conservatively assuming cellulose is degraded. To conduct a safety evaluation, an appropriate degradation model should be applied to determine the degradation rate of cellulose. According to overseas research, despite the mid-chain scission being yet to be seen in the experiments, the degradation model considering mid-chain scission is applied, resulting in an almost 100% degradation rate. The model is selected because the repositories are backfilled with cement, achieving a pH greater than 13, so extensive degradation is reasonably conservative. However, under the domestic disposal condition, where cellulose degradation is unlikely to occur, applying such model would be excessively conservative. Thus, the peeling and stopping model derived by Van Loon and Haas, which suggests 10~25% degradation rate, is reasonably conservative. Based on this model, cellulose would not be a major concern in the domestic 1st phase repository. In the future, this study could be used as fundamental data for planning waste acceptance criteria.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

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.