Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

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.

Porous ceramics are used in various industrial applications based on their physical properties, including isolation, storage, and thermal barrier properties. However, traditional manufacturing environments require additional steps to control artificial pores and limit deformities, because they rely on limited molding methods. To overcome this drawback, many studies have recently focused on fabricating porous structures using additive manufacturing techniques. In particular, the binder jet technology enables high porosity and various types of designs, and avoids the limitations of existing manufacturing processes. In this study, we investigated process optimization for manufacturing porous ceramic filters using the binder jet technology. In binder jet technology, the flowability of the powder used as the base material is an important factor, as well as compatibility with the binder in the process and for the final print. Flow agents and secondary binders were used to optimize the flowability and compatibility of the powders. In addition, the effects of the amount of added glass frit, and changes in sintering temperature on the microstructure, porosity and mechanical properties of the final printed product were investigated.

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.

Background: Individuals with scapular winging have a weak serratus anterior (SA) muscle, and to compensate, the pectoralis major (PM) and upper trapezius (UT) muscles excessively activate, which can cause upper extremity dysfunction. This study aimed to compare the effects of isometric horizontal abduction (IHA) on SA, PM, and UT muscle activity, as well as the SA/PM and SA/UT muscle activity ratios during knee push-up plus (KPP) at 90° and 120° of shoulder flexion. Objects: This study aimed to compare the effects of IHA on SA, PM, and UT muscle activity, as well as the SA/PM and SA/UT muscle activity ratios during KPP at 90° and 120° of shoulder flexion. Methods: This study, conducted at a university research laboratory, included 20 individuals with scapular winging. Participants performed KPP with and without IHA at 90° (KPP90) and 120° (KPP120) of shoulder flexion. SA, PM, and UT muscle activity were measured using surface electromyography. Results: PM activity in KPP90 with IHA was significantly lower than KPP90 and in KPP120 was significantly lower than KPP90. UT activity was significantly greater with IHA than without IHA and at 120° than 90° of shoulder flexion. SA/PM muscle activity ratio was significantly higher in KPP90 with IHA than without IHA and in KPP120 than in KPP90. SA/UT muscle activity ratio was significantly lower with IHA than without IHA. Conclusion: KPP90 with IHA and KPP120 are effective exercises to reduce PM activity and increase SA/PM muscle activity ratio. However, applying IHA in KPP90 also reduces SA/UT muscle activity ratio, implying that it would be preferable to apply KPP120 in individuals overusing their UT muscles.

Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

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.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

Porous ceramics have the advantages of low density, low thermal conductivity, and excellent mechanical properties. Among porous ceramic manufacturing methods, the replica template method allows the easy manufacturing of porous filters with the highest porosity and pores of the desired size, but it also has the disadvantage that the resulting filters have low mechanical strength. To overcome this shortcoming, mullite (3Al2O3·2SiO2) whiskers, which have excellent thermal stability and high mechanical strength, were introduced in porous ceramic structure. The mullite whiskers were synthesized using a composition of Al2O3, flyash and MoO3. The morphologies and crystal structures of the mullite whiskers with MoO3 contents were investigated in detail. When the porous ceramic with mullite whiskers was fabricated using 20 wt% MoO3 catalyst the most uniform microstructure was obtained, and the mullite whiskers showed the highest aspect ratio of 47.03. The porosity and compressive strength of the fabricated porous ceramic were 82.12% and 0.83 MPa, respectively.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

The fabrication of waste forms with high thermal and structural stability is an essential technology for the safe management and disposal of radioactive wastes. In particular, the thermal characteristics of waste forms containing high heat-generating nuclides such as Cs and Sr can be used for the optimized design of the waste form to secure its thermal safety, and they also provide basic design data for the safe management of canisters, storage systems, and repositories. The Korea Atomic Energy Research Institute is actively developing processes and equipment for fabricating various types of high-level wastes into a stable glass or ceramic waste form. In previous research related to the thermal analysis of the waste form, a relatively simple analysis was performed by using the analytic solution of the one-dimensional steady-state heat conduction equation considering the decay heat properties of the waste. As a specific application study, the optimized diameter of the cylindrical glass waste form was proposed by evaluating the centerline temperature of the waste form. In this study, we extended previous research by introducing a more complicated model, and the main results are summarized as follows. First, an analytical solution was derived by applying the temperaturedependent thermal conductivity expressed in the general form of polynomial function to the onedimensional heat conduction problem previously studied. Second, the two-dimensional axisymmetric steady-state heat conduction problem with a more realistic cylinder model with finite length was modeled and solved by using the finite element method via Matlab’s PDE (partial differential equation) toolbox. Third, thermal analysis was performed on the SrTiO3 waste form, selected as a stable form of strontium nuclide, using the developed analytical and numerical methods. The differences in the temperature distribution and computation time were evaluated through a comparative study of both solutions. Although the problem considered in this study could easily be solved by using commercial CFD software such as ANSYS or SolidWorks, a code-based program was developed to facilitate parametric design study in conjunction with optimization algorithms. The analysis results could be used to evaluate the thermal stability of waste form and to optimize the shape and size of the waste form in consideration of the design constraints of storage systems or repositories.

빈번한 가뭄의 피해는 수자원의 양극화를 초래하고 있다. 국내에서는 2011년부터 2017년 사이에 국지적이고 주 기적인 이상 가뭄이 지속되어 10년 가뭄 빈도로 설계된 소규모 저수지의 저수율이 감소하였다. 이러한 저수율의 감소 는 수질을 떨어트릴 수 있고 이미 확보된 용수의 사용마저 제한할 수 있다. 따라서 가뭄에 대비하고 극복하기 위해서 는 수질관리와 가뭄 빈도의 재산정이 필요하다. 이 연구는 저수지에서 수질 변화의 원인인 잠재적 오염물질을 추정하고, 가뭄 기간 동안 저수지의 저수율 감소와 그에 따른 수질 변화를 검토한 것이다.