Protoplasts were isolated from the primary leaves of lettuce (Lactuca sativa L.) seedlings 10 days after in vitro germination. The leaves were stripped and incubated in an enzyme mixture consisting of 1.2% Cellulase R-10 and 0.3% Macerozyme R-10 in cell and protoplast washing solution (CPW) overnight. The average protoplast yield was 8.25 x 106 protoplasts per g of fresh leaf tissue. When protoplasts were cultured at a density of 3.0 × 105 protoplasts/mL in agarose solid KM8P/KM8 medium, first and second divisions were observed in the protoplasts within a week. Protoplast-derived microcolonies formed after 4 weeks of culture, and visible colonies were present after 3 months of culture. Protoplast-derived microcalli were transferred to Murashige and Skoog medium supplemented with 2.0 mg/L kinetin and 0.1 mg/L NAA and incubated in the light for 3 weeks. They grew into callus, which then regenerated into plants after 7 weeks of culture. The regenerated plants grew as apparently normal flowering fertile plants.

담낭암은 드물지만 치명적인 악성 종양으로 초기에는 증상이 없어 대부분의 담낭암이 늦게 진단되고, 급속하게 인접 장기로 전이되기 때문에 예후가 불량하다. 따라서 담낭암을 조기에 발견하는 경우는 드물고, 담낭담석이나 용종으로 수술적 절제를 시행한 후 우연히 발견되는 때가 대부분이다. 저자들은 급성 무결석 담낭염 환자에서 초음파 내시경을 통해 비교적 조기에 담낭관암을 발견하여 성공적으로 치료할 수 있었던 증례를 경험하였다. 이에 문헌고찰과 함께 보고한다.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

The dimensioning machine installed in the hot cell has been used for 20 years. It has been used for a long time so it was often malfunction due to aging and radiation. In addition, some parts of apparatus were discontinued and there were a lot of problems in maintenance and repair. In the old measuring system, the operator’s subjectivity was much involved. The process of control the focal length (distance between lens and specimen) by operator’s sense is a good example. The existing dimensioning machine was the Kh-7700 of Hirox Co., Ltd. As the equipment had been used for a longtime, additional utilities such as jigs, lighting module and servo motors have been customized and used. The same company’s apparatus was selected for the reasons that it did not need to manufacture a new utility so it could save the cost of radioactive waste disposal for existing utilities and its radiation resistance which has been used for 20 years in radiation environment. Lighting, standing, stage, controllers, cables and so on had been customized to provide remote control in hot cell. The installation was completed in March of this year in hot cell and has been successfully used until now. Through the improvement of dimensioning machine, an auto-focusing and multi-focusing were available. Therefore, they made it possible to produce high quality data and improve the accuracy of data. And this research could be a good example of how hot cell devices can be built and customized to achieve remote control.

This study presents an example of creating and optimizing a task sequence required in an automated remote dismantling system using a digital manufacturing system. An automated remote dismantling system using a robotic arm has recently been widely studied to improve the efficiency and safety of the dismantling operations. The task sequence must be verified in advance through discrete eventbased process simulation in a digital manufacturing system to avoid problems in actual remote cutting operations as the main input of the automated remote dismantling system. The laser cutting method can precisely cut complicated target structures such as reactor internals with versatility, but a robot and a pre-prepared program are required to deploy sophisticated motion of the laser cutting head on the target structure. For safe and efficient dismantling operations, the robot’s program must be verified in advance in a virtual environment that can represent the actual dismantling site. This study presents creating and optimizing the task sequence of a robotic underwater laser cutting as part of the project of developing an automated remote dismantling system. A task sequence is created to implement the desired cutting path for the target structure using the automated remote dismantling system in the virtual environment. The task sequence is optimized for the posture of the laser cutting head and the robot to avoid collisions during the operation through discrete event-based process simulation since the target structure is complicated and the volume occupied by the laser cutting head and the robot arm is considerably large. The task sequence verified in the digital manufacturing system is demonstrated by experiments cutting the target structure along the desired cutting path without any problems. The various simulation cases presented in this study are expected to contribute not only to the development of the automated remote dismantling system, but also to the establishment of a safe and efficient dismantling process in the nuclear facility decommissioning.

In the pilot scale test, the two scale-up factors (Electric energy per order EEO, Electric energy per mass EEM) were conducted to design the Chemical Waste Decomposition & Treatment System (CWDS). The CWDS consist of two kind UV lamp reactors to improve the decomposition rate of oxalic acid, which are low pressure amalgam UV lamp and medium pressure UV lamp. The two reactors were connected in series, and the hydrogen peroxide is mixed through a line mixer at the front of the reactor and injected into the reactors. The CWDS was connected with the full system decontamination equipment to purify the residual oxalic acid after chemical decontamination process. The full system decontamination equipment were included Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS) to operate the Oxidation/Reduction decontamination process and purify the process water. After decontamination process, the waste water will be cooled down into the 40°C and passed through the UV reactor at 110 gpm with hydrogen peroxide injection. The concentration of waste water is expected oxalic acid 1,700 ~ 2,000 ppm, Iron 5 ~ 20 ppm. As a result of the CBD test in the laboratory with simulated waste liquid, the amount of Low pressure amalgam lamp UV dose required to decompose 95% of oxalic acid in 2 m2 waste water was up to 1,800 mJ/cm2. The amount of medium pressure lamp UV dose was up to 450 mJ/cm2 at the same condition. We conducted demonstration test using 2 m2 waste water after the oxidation/reduction decontamination process, the decomposition rate 95% was obtained by low pressure amalgam UV lamp and medium pressure UV lamp reactor each.

The radioactive Sr-90, which is formed from beta decay, is well known as one of the most commonly detected nuclides in radioactive waste. In 2015, it was reported that Sr-90 was observed in some soil and metal wastes among the 516 drums of radioactive waste transferred from the decommissioning site of the Korea Research Reactor (in Seoul) to the disposal site (in Gyeongju). Decontamination and sequestration of radionuclides, including Sr, from nuclear waste is important because they are hazardous and harmful to the ecological environment. Immobilization of these nuclides using a zeolite framework is suitable and simple method that has been widely studied. Therefore, it is still necessary to continuously explore the thermal stability of various zeolites and environmental changes around adsorbed cations in zeolite pore for effective immobilization of these radionuclides. In this study, we observed the thermal stability in fully Sr-exchanged natrolite (Sr-NAT), one of small-pore zeolite, from room temperature to 350°C using the in-situ synchrotron X-ray powder diffraction and thermogravimetric (TGA) analysis. In addition, we investigated the structural changes in Sr-NAT during temperature increase by Rietveld analysis. Sr-NAT exhibited apparent zero thermal expansions (ZTE) with the thermal expansion coefficients of -3(1) × 10-6 at the initial stage of increasing the temperature due to dehydration process. In the section from 250°C to 300°C, a phenomenon like negative thermal expansion (NTE) occurs in which the unit cell volume of Sr-NAT decreases despite the increase in temperature. Sr-NAT maintained well its crystallinity up to 350°C, and it became amorphous at 350°C. In this study, we provide a fundamental understanding of the structural changes and thermal stability mechanism of Sr-exchaged zeolite natrolite with increasing temperature.

3D imaging equipment is essential for automated robotic operations that cut radiologically contaminated structure and transfer segmented pieces in nuclear facility dismantling site. Automated dismantling operations using programmed robotic arms can make conventional nuclear facility dismantling operations much more efficient and safer, so dismantling technologies using robotic arms are being actively researched. Resolving the position uncertainty of the target structure is very important in automated robot work, and in general industries, the problem of position uncertainty is solved through the method of teaching the robot in the field, but at the nuclear facility dismantling site, the teaching method by workers is impossible due to activated target structures. Therefore, 3D imaging equipment is a key technology for a remote dismantling system using automated robotic arms at nuclear facility dismantling site where teaching methods are impossible. 3D imaging equipment available in radioactive and underwater environments is required to be developed for a remote dismantling system using robotic arms because most commercial 3D scanners are available in air and certain 3D scanners available in radioactive and underwater environments cannot satisfy requirements of the remote dismantling system such as measurement range and radiation resistance performance. The 3D imaging equipment in this study is developed based on an industrial 3D scanner available in air for efficient development. To protect the industrial 3D scanner against water and radiation, a housing is designed by using mirrors, windows and shieldings. To correct measurement errors caused by refraction, refraction model for the developed 3D imaging equipment is defined and parameter studies for uncertain variables are performed. The 3D imaging equipment based on the industrial 3D scanner has been successfully developed to satisfy the requirements of the remote dismantling system. The 3D imaging equipment can survive up to a cumulative dose of 1 kGy and can measure a 3D point cloud in the air and in water with an error of less than 1 mm. To achieve the requirements, a proper industrial 3D scanner is selected, a housing and shielding for water and radiation protection is designed, refraction correction are performed. The developed 3D imaging equipment is expected to contribute to the wider application of automated robotic operations in radioactive or underwater environments.

The dismantlement of the Kori Unit 1 and Wolsong Unit 1 nuclear power plants is scheduled. Since about 40% of the cost of dismantling nuclear power plants is the cost of disposing of generated wastes, it is important to secure recycling technologies. Among them, low and intermediate level radioactive wastes are made of porous filters and adsorbent materials of ceramic foam to remove nuclides such as C-14, I, and Xe generated during nuclear dismantling. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. Therefore, in this study, the foam filter performance was improved by applying a sacrificial material to increase the specific surface area and porosity of the ceramic foam filter. The sacrificial material is burned out with polyurethane (PU) of the green filter before the heat treatment temperature to increase the strength of the ceramic foam filter so that it can be maintained as pores, thereby improving the specific surface area and porosity. The sacrificial materials and melting temperature (Tm) reviewed in this study were anthracite (530~660°C), PMMA (160°C), Cellulose acetate (260~270°C), and aluminum particle (660°C), and their effect on the manufacture of foam filters was studied by applying this. The specific surface part and porosity of the foam filter were improved when anthracite and aluminum particle were added, and PMMA and Cellulose acetate, which are relatively low temperature melting points, were burned out at a temperature lower than PU, and thus their physical properties were not greatly affected. The physical properties and specific surface part and porosity of ceramic foam filters manufactured using various sacrificial materials will be discussed.

Korea Radioactive Waste Agency (KORAD), regulatory body and civic groups are calling for an infrastructure system that can more systematically and safely manage data on the results of radioactive waste sampling and nuclide analysis in accordance with radioactive waste disposal standards. To solve this problem, a study has been conducted on the analysis of the nuclide pattern of radioactive waste on the nuclide data contained in low-and intermediate-level radioactive waste. This paper will explain the optimal repackaged algorithm for reducing radioactive waste based on previous research results. The optimal repackaged algorithm for radioactive waste reduction is comprised based on nuclide pattern association indicators, classification by nuclide level of small-packaged waste, and nuclide concentration. Optimization simulation is carried out in the order of deriving nuclide concentration by small-packaged, normalizing drum minimization as a function of purpose, normalizing constraints, and optimization. Two scenarios were applied to the simulation. In Scenario 1 (generating facilities and repackaged by medium classification without optimization), it was assumed that there are 886 low-level drums and 52 very low-level drums. In Scenario 2 (generating facilities and repackaged by medium classification with optimization), 708 and 230 drums were assigned to the low-level and very low-level drums, respectively. As a result of the simulation, when repackaged in consideration of the nuclide concentration and constraints according to the generating facility cluster & middle classification by small package (Scenario 2) the low-level drum had the effect of reducing 178 drums from the baseline value of 886 drums to 708 drums. It was found that the reduced packages were moved to the very low-level drum. The system that manages the full life-cycle of radioactive waste can be operated effectively only when the function of predicting or tracking the occurrence of radioactive waste drums from the source of radioactive waste to the disposal site is secured. If the main factors affecting the concentration and pattern of nuclides are systematically managed through these systems, the system will be used as a useful tool for policy decisions that can prevent human error and drastically reduce the generation of disposable drums.

According to the continued generation of spent nuclear fuel, a reliable safety assessment is highly required with the precise modeling of the migration and retardation behavior of radionuclides to enhance public acceptance and hinder excessive conservativeness during the construction of the repository. In particular, the colloids formed in the repository-relevant condition are known to accelerate the migration of radionuclides. Thus, geochemical behavior and relevant characteristics of colloids are needed to be unambiguously clarified. The objective of the present work is to investigate the fundamental characteristics of colloids contained in the natural groundwater system by using various analytical methods and the tangential flow ultra-filtration (TFUF) system. The granitic groundwater sample from the DB-3 borehole at the KURT (KAERI Underground Research Tunnel) was taken by an airtight stainless steel cylinder coated on the inside with PTFE to prevent the infiltration of ambient air into the geologic groundwater sample. And then, the groundwater sample was transferred to the inert glovebox filled with Ar gas to monitor the pH and Eh equilibrium of the aqueous sample. For further investigation, the colloid contained in the groundwater sample was concentrated by using the TFUF system equipped with a membrane filter (pore size: 3 kDa). The concentrated groundwater sample was analyzed with various methods such as ICP-MS/OES, IC, DLS/ELS, FE-TEM/SEM-EDS, ATR-FTIR, TOC, LC-OCD, etc. In this study, the size of groundwater colloids was determined to be 182.3 ± 52.7 nm with the major constituents of C, S, O, Fe, Al, Si, etc. The amount of organic carbon and the concentrations of organic substances determined by means of the molecular weight fraction with the TOC and LC-OCD provide further detailed information for the colloids in the KURT groundwater sample. The results obtained in this study are expected to be used as preliminary experimental data for modeling the colloid-facilitated migration of radionuclides to improve the reliability of the safety assessment of the geologic repository.