Various types of optical materials and devices used in special environments must satisfy durability and optical properties. In order to improve the durability of zinc sulfide multispectral (MS ZnS) substrates with transmission wavelengths from visible to infrared, Ge-Sb-Se-based chalcogenide glass was used as a sealing material to bond the MS ZnS substrates. Wetting tests of the Ge-Sb-Se-based chalcogenide glass were conducted to analyze flowability as a function of temperature, by considering the glass transition temperature (Tg) and softening temperature (Ts). In the wetting test, the viscous flow of the chalcogenide glass sample was analyzed according to the temperature. After placing the chalcogenide glass disk between MS ZnS substrates (20 × 30 mm), the sealing test was performed at a temperature of 485 °C for 60 min. Notably, it was found that the Ge-Sb-Se-based chalcogenide glass sealed the MS ZnS substrates well. After the MS ZnS substrates were sealed with chalcogenide glass, they showed a transmission of 55 % over 3~12 μm. The tensile strength of the sealed MS ZnS substrates with Ge-Sb-Se-based chalcogenide glass was analyzed by applying a maximum load of about 240 N, confirming its suitability as a sealing material in the far infrared range.

The ambiguity of complex regional pain syndrome (CRPS) posed a challenge to many medical researchers in the early days after its discovery and continues to do so till date. The establishment of the Budapest Criteria of the International Association for the Study of Pain resolved certain queries on CRPS. Many aspects of CRPS, such as pathophysiology and etiology, remain unknown. Therefore, of these aspects, we focused on the genetic basis of CRPS. In this qualitative review, we summarized the recent findings on the genetic association of CRPS and analyzed the roles of genes identified in each study and limitations of the studies. In particular, we confirmed the reliability of each study by comparing the following research, which used the following control groups or the same candidate genes. Notably, specific phenotypes of CRPS with dystonia indicate a significant association with human leukocyte antigen (HLA)-DQ8. Further, HLA-DQ8, which is associated with aberrant CD4+ T-cell reaction, could be associated with CRPS etiology since an increased CD4+ T-cell population was reported in CRPS patients. In addition, matrix metalloproteinase (MMP)-9 found in genome-wide expression profiling is noteworthy since MMP-9 is associated with neuro-inflammatory reactions. Despite these suggestions on the genetic aspects of CRPS, the pathophysiology and etiology of CRPS may be polygenic and multifactorial, influenced by multiple genes and other factors. Further, some studies have suggested that CRPS phenotypes have different etiologies. Thus, further studies with the precise classification of CRPS on a unified basis and with a significant number of case groups are required

North Korea continues to upgrade and display its long-range rocket launchers to emphasize its military strength. Recently Republic of Korea kicked off the development of anti-artillery interception system similar to Israel’s “Iron Dome”, designed to protect against North Korea’s arsenal of long-range rockets. The system may not work smoothly without the function assigning interceptors to incoming various-caliber artillery rockets. We view the assignment task as a dynamic weapon target assignment (DWTA) problem. DWTA is a multistage decision process in which decision in a stage affects decision processes and its results in the subsequent stages. We represent the DWTA problem as a Markov decision process (MDP). Distance from Seoul to North Korea’s multiple rocket launchers positioned near the border, limits the processing time of the model solver within only a few second. It is impossible to compute the exact optimal solution within the allowed time interval due to the curse of dimensionality inherently in MDP model of practical DWTA problem. We apply two reinforcement-based algorithms to get the approximate solution of the MDP model within the time limit. To check the quality of the approximate solution, we adopt Shoot-Shoot-Look(SSL) policy as a baseline. Simulation results showed that both algorithms provide better solution than the solution from the baseline strategy.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.

Se-79, a fission product of uranium, is present in spent nuclear fuel. Selenium is volatilized from the spent nuclear fuel during the pretreatment of pyroprocessing, and a filter composed of calcium oxide can capture gaseous selenium in the form of CaSeO3. Because Se-79 has a long half-life (3.27E5 years) and selenite ions have high mobility in groundwater, they must be immobilized in a chemically stable form for final disposal. This study used a composition of 50 TeO2 - 10 Al2O3 - 10 B2O3 - 10 Na2O - 10 CaO - 10 ZnO (mol%). High-purity powders of TeO2, Al2O3, H3BO3, Na2CO3, CaCO3, and ZnO were used as glass precursors. The mixed powders were placed in alumina crucibles and melted in an electric furnace under an ambient atmosphere at 800°C for 1 h before being cast on a carbon mold. The obtained glasses were ground into fine powders and then mixed with CaSeO3 powders. The powders were melted in alumina crucibles at 800°C for 1 h. To simulate a seleniumcaptured calcium filter, CaSeO3 was synthesized by a precipitation method using sodium selenite (Na2SeO3) and calcium nitrate (Ca(NO3)2) solutions. The glass samples were analyzed by an X-ray diffractometer (XRD). Retention of Se in tellurite glasses was analyzed by an X-ray fluorescence spectrometer (XRF) and inductively coupled plasma (ICP). The chemical durability of tellurite glass was evaluated through the PCT method.

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.

The number of nuclear power plants that are permanently shut down or decommissioned is increasing worldwide, and accordingly, research is being conducted on an appropriate method for disposing of radioactive waste generated during the decommissioning of nuclear power plants. In the case of waste liquid generated during the decommissioning of nuclear power plants, it is important not only to efficiently reduce waste but also to secure the suitability of disposal. One of the solidification treatment methods for radioactive waste is cement solidification, but since cement solidification has poor solidification properties and generates a large amount of waste, improvement activities have been pursued. This study aims to develop high-performance cement-based materials and solidification treatment technology for solidification of liquid radioactive waste generated during nuclear decommissioning in order to improve the problems of cement solidification treatment method. For the development of polymer cement, epoxy resin and polyamine/amide mixed type and general Portland cement were mixed in various ratios. The most appropriate mixing ratio was 4.5:2, which showed the highest compressive strength. A simulated waste liquid was prepared by referring to the preliminary decommissioning plan of Shin-Kori Units 5 and 6, and it was dried and made into granules. Polymer cement was injected into a drum filled with granules by vacuum pressure to prepare a waste form matrix. In the solidification process, granules made by drying the waste liquid were used, and the solidification agent was filled in between the granules, so the total volume of solid radwaste was reduced compared to the conventional cement solidification treatment method. As a result, the amount of waste decreased to about 1/3, and the volume reduction rate increased by about 2.2 times. The compressive strength of 3,243 psi was confirmed in the disposability performance test for the manufactured solid samples. The compressive strength after the thermal cycling test, irradiation test, microorganism test, and immersion test was 2,257 psi, 2,306 psi, 4,530 psi, and 2,263 psi, respectively, exceeding the acceptance criteria of 500 psi. The leaching index was 7~13, and no free standing water was generated.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

Spent nuclear fuel (SNF) is the main source of high-level radioactive wastes (HLWs), which contains approximately 96% of uranium (U). For the safe disposal of the HLWs, the SNF is packed in canisters of cast iron and copper, and then is emplaced within 500 m of host rock surrounded by compacted bentonite clay buffer for at least 100,000 years. However, in case of the failure of the multi-barrier disposal system, U might be migrated through the rock fractures and groundwater, eventually, it could reach to the biosphere. Since the dissolved U interacts with indigenous bacteria under natural and engineered environments over the long storage periods of geologic disposal, it is important to understand the characteristics of U-microbe interactions under the geochemical conditions. In particular, a few of bacteria, i.e., sulfate-reducing bacteria (SRB), are able to reduce soluble U(VI) into insoluble U(IV) under anaerobic conditions by using their metabolisms, resulting in the immobilization of U. In this study, the aqueous U(VI) removal performance and change in bacterial community in response to the indigenous bacteria were investigated to understand the interactions of U-microbe under anaerobic conditions. Three different indigenous bacteria obtained from different depths of granitic groundwater (S1: 44–60 m, S2: 92–116 m, and S3: 234–244 m) were used for the reduction of U(VI)aq. After the anaerobic reaction of 24 weeks, the changes in bacterial community structure in response to the seeding indigenous bacteria were observed by high-throughput 16S rDNA gene sequencing analysis. The highest uranium removal efficiency of 57.8% was obtained in S3 sample, and followed by S2 (43.1%) and S1 (37.7%). Interestingly, SRB capable of reducing U(VI) greatly increased from 4.8% to 44.1% in S3 sample. Among the SRB identified, Thermodesulfovibrio yellowstonii played a key role on the removal of U(VI)aq. Transmission electron microscopy (TEM) analysis showed that the dspacing of precipitates observed in this study was identical with that of uraninite (UO2). This study presents the potential of U(VI)aq removal by indigenous bacteria under deep geological environment.