Pleurotus ostreatus is a globally cultivated mushroom crop. Cap color is a quality factor in P. ostreatus. However, cap color can spontaneously mutate, degrading the quality of the mushroom on the market. Early detection and removal of mutant strains is the best way to maintain the commercial value of the crop. To detect the cap color mutant Gonji7ho, molecular markers were developed based on insertion/deletions (InDels) derived from the comparison of mitogenomes of Gonji7ho and Gonji7hoM mushrooms. Sequencing, assembly, and comparative analysis of the two mitogenomes revealed genome sizes of 73,212 bp and 72,576 bp with 61 and 57 genes or open reading frames (ORFs) in P. ostreatus Gonji7ho and Gonji7hoM, respectively. Fourteen core protein-encoding genes, two rRNA, and 24 tRNA with some OFRs were predicted. Of the 61 genes or OFRs in the wild type, dpo, rpo, and two orf139 were missing (or remnant) in the mutant strain. Molecular markers were developed based on the sequence variations (InDels) between the two mitogenomes. Six polymorphic molecular markers could detect the mutated mitochondria by PCR. These results provide basic knowledge of the mitogenomes of wild-type and mutant P. ostreatus, and can be applied to discriminate mutated mitochondria.

Veronica L., the largest genus in the family Plantaginaceae, is widespread in various habitats. Due to their long-blooming flowers, Veronica species have high horticultural value as indoor potted, garden, and landscape plants. Furthermore, Veronica plants are extremely important owing to their notable diversity in habitat usage, ploidy level, and evolution. Several native taxa, which are of key interest in breeding programs and phylogenetic studies, have been identified in Korea. The genome sizes and chromosomal characteristics are basic cytogenetic features of all taxa, and their knowledge is a prerequisite when commencing genome sequencing projects. It can provide essential information for cytogenetic, taxonomic, phylogenetic, and evolutionary studies. Thus, cytogenetic analysis and genome size estimation of seven Veronica taxa native to Korea were conducted in this study. Fluorescence in situ hybridization (FISH) karyotype analysis and chromosome counting was conducted using metaphase chromosomes probed with 5S and 45S rDNA. Nuclear DNA content and genome size were determined using flow cytometry. FISH karyotype analysis revealed a common number of 5S loci and varying 45S signals that create distinctive rDNA distribution patterns in each taxon. The results indicated that the seven investigated Veronica taxa have calculated genome sizes (1C values) ranging from 517.1 to 862.0 Mbp. This study is the first to report the chromosome number and karyomorphology of seven Veronica taxa native to Korea, as well as the use of rDNA markers for identifying individual chromosomes. These findings contribute to the crucial understanding the genomic characteristics of species within the genus Veronica, serve as a basis for studying Veronica phylogeny and evolution, and provide valuable information for future breeding programs.

At high temperatures, molten salt has heat transfer properties like water. Molten salt has the characteristics of a strong natural circulation tendency, large heat capacity, and low thermal conductivity. Unlike sodium, molten salt does not react explosively exothermically with air. However, molten salt has a strong tendency to corrode materials, and its properties are easily changed by a sensitive reaction to oxygen and moisture. Therefore, it is necessary to study material corrosion properties and chemical control methods for nuclear fuel salts, which are eutectic mixtures. In this study, the optimal operation method of the thermal convection loop is established to perform the experiments on the molten salt. The process describes briefly as follows. The operation step consists of preparation, purification, transportation, and operation. In the preparation, the step checks the entire structure and equipment (TC, blower, vacuum pump, etc.). And melt the salt mixture at a high temperature (670°C) slowly in the purification step. Before injecting the molten salt, the surface temperature of the entire loop must retain temperature (about 500°C) constantly. Completely melted molten salt in the melting pot is flow along the pipe of the thermal convection loop in the transportation step. Lastly, the convection of molten salt goes to keep by the temperature difference. The thermal convection loop can be utilized for various experiments such as corrosion tests, component analyses, chemistry control, etc.

Molten salt used in the multipurpose molten salt experiment must be of high purity. Depending on the purpose of the experiment, only the base component of the molten salt be used, or a component simulating a nuclear fission product be added to the base component and used. In all cases, an increase in the concentration of impurities such as oxygen and moisture may lead to an erroneous interpretation when analyzing the experimental results. Therefore, molten salt should be purified before use. In this study, the purification of molten salt is described for multi-purpose molten salt experiments. The salt mixture is selected as MgCl2-NaCl and is quantified at a mixing ratio of 43mol%:57mol%. The salt mixture is treated in a glove box environment because of must minimize the reaction of adsorbing oxygen and moisture when the salt mixture is exposed to the atmosphere. MgCl2 is more likely to contain water than NaCl, the purification of the NaCl-MgCl2 mixture is established according to the purification process for removing water from MgCl2. A process for purifying the salt mixture briefly consists as follows: drying moisture, melting salts, purification, removing HCl, and stabilization. Through the process be able to obtain high-purity molten salt and more accurate experiment results.

As the number of aging nuclear power plants increases, the market for dismantling nuclear power plants is growing rapidly. About 40% of the cost of dismantling nuclear power plants is the waste treatment cost incurred during the dismantling process, of which concrete waste accounts for a significant portion of the total waste. Securing technology for reducing and recycling concrete waste is very important not only in terms of economy but also in terms of environment. The objective is to synthesize geopolymer using inorganic materials from cement fine powder in concrete waste. Cement fine powder in concrete waste has a large amount of inorganic elements necessary for filing materials for radioactive waste treatment such as CaO and SiO2. In particular, Ca(OH)2 is synthesized by extracting Ca2+ from concrete waste. It can be used as an alkali activator to synthesize geopolymer. The mortar from crushed concrete was used as a source of calcium. The first step is to react with concrete waste and hydrochloric acid to extract ions. The second step is to react with NaOH and synthesize Ca(OH)2. The product was divided into two stages according to the reaction method and order. The first and second products were washed and dried, and then XRD and XRF were performed. The second product was matched only Ca(OH)2 and CaCO3 at the XRD peak. In the case of XRF, it was analyzed to have a purity of 67.80–78.73%. Synthesis of geopolymer by recycling materials extracted from concrete waste can reduce disposal costs and improve the utilization rate of disposal sites.

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.

During the treatment of spent nuclear fuel, radioactive iodine is generated in a liquefied or gaseous form in a specific process. In the case of iodine 129, it is a long-lived nuclide with a very long halflife and has high groundwater mobility under repository conditions. Despite showing a low radioactivity value, research on the management of radioactive iodine from a long-term perspective is continuously being performed. Although research has been conducted using borosilicate glass as a medium for solidifying iodine, compatibility of I in borosilicate glass is very small and the volatility is high in the solidification process. So it is not suitable as a solidified substance of iodine. Therefore, studies on other solidification media to replace them are continuously being conducted. Our research team tried to develop a new medium that can contain iodine in a solidified body stably through a simple heat treatment process and can improve problems such as volatility and waste loading. Iodine is captured as AgI in the Ag ion-exchanged zeolite. So, TeO2, Ag2O, and Bi2O3 having a high AgI loading rate were used as main components. It was named TAB after taking the first letter of each element. In previous studies, the physical properties, structure, and chemical stability of TAB materials were confirmed. PCT (Product Consistent test) was performed to confirm chemical stability. It is mainly used to compare the chemical stability of glass materials with other glass materials, but there are limitations in evaluating the long-term chemical stability of materials. In this experiment, we tried to evaluate the long-term stability of TAB and compare it with borosilicate, which is conventionally used to treat radioactive waste. In addition, we tried to understand the leaching behavior inside the TAB medium. For this purpose, ASTM C1308 test was performed for 365 days, and distilled water and KURT groundwater were used as leachates to examine the effect of ions in the groundwater on the solidified body. To analyze the leaching behavior, ICP-MS and ICP-OES analyses were performed, and the cross-section of the sample after leaching was observed through SEM.

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.