Chronic fibroinflammatory pancreatitis causes irreparable damage to the pancreatic parenchyma. This frequently results in food restrictions, painkiller addiction, and serious quality of life impairment in children. We observed a 13-year-old girl who had previously been hospitalized multiple times and had undergone acute pancreatitis arrived with excruciating abdominal pain and recurrent hematemesis. A chronic intra-pseudocyst bleed and an ampulla of Vater hemorrhage were discovered during an upper gastrointestinal endoscopy. A 3×2×1 cm pancreatic head pseudocyst was discovered on the magnetic resonance cholangiopancreatography, however the computed tomography scan revealed a pancreatic head pseudocyst, pancreatic duct stones, and substantial peripancreatic inflammation. The multidisciplinary team determined that Frey’s surgery was the most efficient method to minimize her suffering. No complications occurred during the healing phase following surgery, and two years later, neither recurrence bleeding nor abdominal pain appeared. In summary, Frey’s method is a secure and efficient intervention when applied by a multidisciplinary team.

We present a new method which constructs an Hi super-profile of a galaxy which is based on profile decomposition analysis. The decomposed velocity profiles of an Hi data cube with an optimal number of Gaussian components are co-added after being aligned in velocity with respect to their centroid velocities. This is compared to the previous approach where no prior profile decomposition is made for the velocity profiles being stacked. The S/N improved super-profile is useful for deriving the galaxy’s global Hi properties like velocity dispersion and mass from observations which do not provide sufficient surface brightness sensitivity for the galaxy. As a practical test, we apply our new method to 64 high-resolution Hi data cubes of nearby galaxies in the local Universe which are taken from THINGS and LITTLE THINGS. In addition, we also construct two additional Hi super-profiles of the sample galaxies using symmetric and all velocity profiles of the cubes whose centroid velocities are determined from Hermite h3 polynomial fitting, respectively. We find that the Hi super-profiles constructed using the new method have narrower cores and broader wings in shape than the other two super-profiles. This is mainly due to the effect of either asymmetric velocity profiles’ central velocity bias or the removal of asymmetric velocity profiles in the previous methods on the resulting Hi super-profiles. We discuss how the shapes (σn/σb, An/Ab, and An/Atot) of the new Hi super-profiles which are measured from a double Gaussian fit are correlated with star formation rates of the sample galaxies and are compared with those of the other two super-profiles.

Anderson-type polyoxometalate (POM) with general formula of [Hy(XO6)M6O18]n- (y=0-6, n=2-8, M=addenda atom, X=heteroatom) represents one of the basic topological structures among POM-type family. Anderson-type POMs have a planar arrangement and two terminal oxygen atoms attached to each addenda metal atom unlike other type. Thus, the Anderson-type POMs have high reactivity and various coordination modes. The various multifunctional organic-inorganic hybrid materials can be synthesized using the Anderson-type POMs as an inorganic building block. Another important feature of the Anderson-type POMs is the incorporation of the heteroatoms with various sizes and oxidation states, which can lead to tune chemical properties. No Anderson-type POMs with early transition metal ions in the heteroatom site have been reported previously. Recently, we reported the synthesis of titanium-containing Anderson-type POM, Na2K6Ti0.92W6.08O24∙12H2O (Ti-POM), which consists of pure inorganic framework built from a central Ti atom surrounded by six WO6 inorganic scaffold. Herein, in-depth studies were conducted to find optimal synthesis conditions such as composition and pH. The success of synthesis was confirmed with Powder X-ray Diffraction that the Ti-POM has a rhombohedral structure with space group of R-3m (No. 166) when the TiOSO4·xH2SO4∙yH2O/ Na2WO4∙2H2O molar ratio is in the range of 0.07 to 0.33. But outside of this range, other unwanted phases coexist. In a basic condition (pH=12), a single-phase Ti-POM with good crystallinity can be obtained, while a Keggin-type POM, NaxK10-x(H2W12O42), was formed through the decomposition of Ti-POM at pH lower than 7.

Attempts to use the molten salt system in various aspects such as MSR or energy storage systems are increasing. However, there are limitations in the molten salt-assisted technique due to the harsh corrosiveness of the molten salt, and a more detailed study on salt-induced corrosion is needed to solve this problem. In this study, corrosion behaviors of 80Ni-20Cr alloy in various salt environments such as eutectic NaCl-MgCl2 with NiCl2, CrCl2, and EuCl3 additives were investigated. Meanwhile, the corrosion acceleration effects of 80Ni-20Cr specimens were analyzed for various ceramic materials such as SiC, Al2O3, SiO2, graphite, and BN, and metallic materials such as Ni-based alloy, Fe-based alloy, and pure metals in a molten salt environment. The experiments were conducted at 973 K for up to 28 days, and after the experiment, the microstructural change of the specimen was analyzed through SEM-EDS, and salt condition was analyzed by ICP-OES.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

The Korea Atomic Energy Research Institute operates the Nuclear Cycle Experimental Research Facility which has radiation controlled area in the laboratory with the aim of realizing pyroprocessing technology. In this Facility, depleted Uranium feed material and a depleted Uranium mixed with some surrogate material are used for performing experiments. Therefore the facility is using uranium, users should be careful of radiation. This paper will explain the radiation protection of the Nuclear Cycle Experimental Research facility and will also explain how much alpha radiation comes out from the facility. The RMS (Radiation Monitoring System) detector is made by CANBERRA and the model name is ICAM. ICAM RMS is the detector which can detect Alpha Radiation by absorbing the air in the facility. The RMS detector is installed in the HVAC room on the third floor to check the air contamination through the chimney. The RMS is connected to the air ventilation line for detecting Alpha radiation in the whole facility. Experiment was performed for two weeks to check the radiation level and the air ventilation fan continued to operate 24 hours a day. the results are below the required value which is 0.1 Bq/m3, indicating that the facility is safe in terms of radiation safety management.

This study presents distribution of naturally occurring radioactive materials in groundwater in Jeju island. Radon (222Rn) and potassium (40K) concentrations were performed by using RAD H2O of RAD7 and 940 Professional IC Vario, respectively. In addition, the activities of uranium and thorium nuclides were analyzed by ICP-MS. All of the groundwater samples were collected from 29 sites from August to October 2022. The radon concentrations in groundwater were in the range of 0 to 60 Bq L-1, and there was no groundwater exceeding the range of 148 Bq L-1 proposed by the US EPA. The distribution of uranium in groundwater varied from 0 to 0.6 μg L-1 and did not exceed 30 μg L-1, thresholds indicated by the US EPA.

Dose-rate monitoring instruments are indispensable to protect workers from the potential risk of radiation exposure, and are commonly calibrated in terms of the ambient dose equivalent (H*(10)), an operational quantity that is widely used for area monitoring. Plastic scintillation detectors are ideal equipment for dosimetry because of their advantages of low cost and tissue equivalence. However, these detectors are rarely used owing to the characteristics caused by low-atomic-number elements, such as low interaction coefficients and poor gamma-ray spectroscopy. In this study, we calculated the G(E) function to utilize a plastic scintillation detector in spectroscopic dosimetry applications. Numerous spectra with arbitrary energies of gamma rays and their H*(10) were calculated using Monte Carlo simulations and were used to obtain the G(E) function. We acquired three different types of G(E) functions using the least-square and first-order methods. The performances of the G(E) functions were compared with one another, including the conventional total counting method. The performance was evaluated using 133Ba, 137Cs, 152Eu, and 60Co radioisotopes in terms of the mean absolute percentage error between the predicted and true H*(10) values. In addition, we confirmed that the dose-rate prediction errors were within acceptable uncertainty ranges and that the energy responses to 137Cs of the G(E) function satisfied the criteria recommended by the International Commission.

Republic of Korea (ROK) is operating the Integrated Environmental Radiation Monitoring Network (IERNet) in preparation for a radioactive emergency based on Article 105 of the Nuclear Safety Act (Monitoring of Nationwide Radioactive Environment). 215 radiation monitoring posts are monitoring a wide area, but their location is fixed, so they can’t cover areas where the post is not equipped around the Nuclear Power Plants (NPPs). For this, a mobile radiation monitoring system was developed using a drone or vehicle. However, there are disadvantages: it is performed only at a specific cycle, and an additional workforce is required. In this study, a radiation monitoring system using public transportation was developed to solve the above problems. Considering the range of dose rates from environmental radiation to high radiation doses in accidents, the detector was designed by combining NaI (TI) (in the low-dose area) and GM detector (in the high-dose area). Field test was conducted by installed on a city bus operated by Yeonggwang-gun to confirm the performance of the radiation monitoring system. As a result of the field test, it was confirmed that data is transmitted from the module to the server program in both directions. Based on this study, it will be possible to improve the radiation monitoring capability near nuclear facilities.

This study evaluated the synthesis of optimal materials for high efficiency adsorption and removal characteristics of Cs-137 for radioactive contaminated water, and considered thermal treatment methods to stabilize the spent adsorbent generated after treatment. We synthesized a composite adsorbent with a combination of impregnating metal ferrocyanide that improves the selectivity of Cs adsorption with zeolite capable of removing Cs as a support. The Cs removal efficiency of the composite adsorbent was evaluated, and the stability change of Cs according to the high-temperature sintering was evaluated as a stabilization method of the spent adsorbent. The metal ferrocyanide content of the adsorbent was in the range of 11.8~36.0%. The adsorption experiments were performed using a simulated liquid waste to have a total Cs concentration of 1 mg/L while containing a trace amount of Cs-137, and then gamma radioactivity was analyzed. In order to evaluate the stabilization of the spent adsorbent, heat treatment was performed in the range of 500~1,100°C, and the volatilization rate of Cs during heat treatment and the leaching rate of Cs after heat treatment were compared. In the adsorption experiment, the Cs removal efficiency was higher than 99%, regardless of the amount of metal ferrocyanide in the composite adsorbent. In the sintering experiment on the spent adsorbent, it was confirmed that there was no volatilization of Cs up to 850°C, and then the volatilization rate increased as the heating temperature increased. On the other hand, the leaching rate of Cs in the sintered adsorbent tends to significantly decrease as the heating temperature increases, so that Cs can be stabilized in the sintered body. In addition, as the content of metal ferrocyanide increases, the volatilization rate of Cs rapidly increases, indicating that the unstable metal ferrocyanide in the adsorbent may adversely affect the removal of Cs as well as the thermal treatment stability.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

To transport radioactive waste generated during the decommissioning of Kori Unit 1, transport containers of various sizes are being developed. Since these radioactive decommissioning waste transport containers are larger than the specifications of the existing IP-2 type transport containers, which are for operational radioactive waste, design of the CHEONG-JEONG-NURI needs to be changed when transporting them to disposal facility using the CHEONG-JEONG-NURI, which carries operational radioactive waste. In this study, design changes of the CHEONG-JEONG-NURI, cargo hold modification plan for efficient loading of radioactive decommissioning waste transport containers and radioactive decommissioning waste container loading arrangement (plan) were evaluated during the design life period (year 2034). First, as only the IP-2 type transport container with a weight of 7.5 tons and size of 1.6 m (W) × 3.4 m (L) × 1.2m (H) can be loaded in the cargo hold, if only the decommissioning radioactive waste containers are to be loaded and transported, cargo hold needs to be reinforced. Second, when both the radioactive decommissioning waste transport container of the same size as the current operating radioactive waste transport container, and the radioactive decommissioning waste transport container of the same size as the ISO-type transport container are to be loaded in the cargo hold of the CHEONG-JEONG-NURI and transported, the overall design changes (cargo hold size and load reinforcement) are required. Third, since the safe working load of the CHEONG-JEONG-NURI crane is 12.5-tons, it shall be replaced with a ship crane of 35-tons or more to handle the decommissioning radioactive waste container smoothly, or a gantry crane used in general port facilities shall be installed. When replacing with a ship crane of 35-tons or more, ship buoyancy, ship stability, and ship structural safety shall be considered. The possibility of moving in all 4 directions for smooth operation, and the possibility of lifting the transport container to a position higher than the height of the CHEONG -JEONG-NURI shall be considered. Loading and transporting all decommissioning radioactive waste containers, which are the same size as IP-2 and ISO-type transport containers, in the cargo hold of the CHEONG-JEONG-NURI is uneconomical due to the need for overall design changes (cargo size and load reinforcement, etc.). Also, delay in delivery of the operation wastes is expected due to a long-term design change period. Therefore, it is considered reasonable to load and transport only the decommissioning radioactive waste transport container, which is the same size as the IP-2 transport container, in the cargo hold.

Reactor pressure vessels and steam generators generated in the process of dismantling nuclear power plants or replaced steam have various shape and occupy a considerable amount of the disposal site when disposed of in original shape. For the development of domestic technologies related to the disposal of large wastes, it is necessary to secure technologies for reducing large radioactive metal wastes, including technologies such as decontamination, cutting, melting, and residual radioactivity evaluation. Cases of disposal of steam generators in Europe and the United States were investigated. Except for u-tubes, steam generators are less contaminated or easily decontaminated, so it is possible to reduce the volume of waste subject to final disposal by exempting a significant amount through appropriate treatment. Korea Hydro & Nuclear Power Co. is currently temporarily storing 24 steam generators at 41.6 billion won. This paper presents a method to exempt more parts of the steam generator and reduce the volume of waste by properly combining mechanical cutting thermal cutting and melting to dispose of the steam generator. Currently the decontamination and dismantling industries of nuclear facilities are gradually expanding around the world. Therefore, it is necessary to localize the treatment technology for metal waste generated during maintenance and dismantling. The result of this study can be used to establish waste reduction and disposal method for dismantling steam generators.

In preparation for the decommissioning of Kori unit 1 of the nuclear power plant (NPP), new containers of package, transportation, and disposal are being developed that reflect the type, generation amount, and radiological characteristics of decommissioning waste. The containers under development have internal volumes of 1 m3 ~ 14 m3 and loading weights of 1 ton ~ 35 tons, which are larger in size and have a higher loadable weight compared to the 200 L drum and IP-2 type transport container currently being used for packaging and transporting waste. So, there is a limit to handling new containers using existing transport systems (cranes, spreaders, forklifts, transport vehicles, etc.). Therefore, in this study, the status of handling equipment in NPP and disposal facilities was reviewed, the flow from the generation to disposal of decommissioning waste was analyzed, and the possibility of handling new container or the necessity of introducing new systems were derived. Except for some high-dose/high-radioactive wastes among decommissioning wastes, all wastes are finally disposed of through decommissioning area, temporary storage facility, waste treatment facility, waste storage facility, and receipt and storage building. The decommissioning area, temporary storage facility, and waste treatment facility are newly established areas for the decommissioning and should be equipped with a spreader/crane with a lifting weight of 15 tons, 35 tons, and 40 tons in consideration of the weight of the package to be handled in the zone. The waste storage facility has a 7.5 tons crane, so it can handle only some of the lower weight of the 5 to 35 tons package that is expected to be handled. Therefore, additional installation of spreaders/cranes, each with a lifting capacity of 15 tons and 40 tons, is required. The maximum loading weight of forklifts and transport vehicles operating at NPP, and disposal facilities is 10 tons and 12.6 tons, respectively. To transport the package, the facility must additionally install 15 tons and 40 tons forklifts, and 40 tons transport vehicles. Since the lifting weight of the crane installed on the transport vessel is also low at 12.5 tons, it is necessary to change the design of the existing or replace it with 40 tons to handle high-weight package. The results of this study will be used as basic data for the establishment of transport systems in the relevant area and facility, and design requirements for each equipment will be derived through additional research.

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

Organic scintillator is easy to manufacture a large size and the fluorescence decay time is short. However, it is not suitable for gamma measurement because it is composed of a low atomic number material. Organic scintillation detectors are widely used to check the presence or absence of radiation. The fluorescence of organic scintillators is produced by transitions between the energy levels of single molecules. In this study, an organic scintillator development study was conducted for use in gamma measurement, alternative materials for secondary solute used in basic organic scintillators were investigated, and the availability of alternative materials, detection characteristics, and neutron/gamma identification tests were performed. In other words, a secondary solute showing an improved energy transfer rate than the existing material was reported, and the performance was evaluated. 7-Diethylamino -4-methylcoumarin (DMC), selected as an alternative material, is a benzopyrone derivative in the form of colorless crystals, has high fluorescence and high quantum yield in the visible region, and has excellent light stability. In addition, it has a large Stokes shift characteristic, and solubility in solvent is good. Through this study, it was analyzed that the absorption wavelength range of DMC coincided with the emission wavelength range of PPO, which is the primary solute. Through this study, it was confirmed that the optimal concentration of DMC was 0.04wt%. As a result of performing gamma and neutron measurement tests using a DMC-based liquid scintillator, it showed good performance (FOM=1.42) compared to a commercial liquid scintillator. Therefore, the possibility of use as a secondary solute was demonstrated. Based on this, if studies on changes in the composition of secondary solute or the use of nanoparticles are conducted, it will be possible to manufacture and utilize a scintillator with improved efficiency compared to the existing scintillator.

The safe, efficient and cost-effective decommissioning and dismantling of radioactive facilities requires the accurate characterization of the radionuclide activities and dose rate environment. And it is critical across many nuclear industries to identify and locate sources of radiation accurately and quickly. One of the more challenging aspects of dealing with radiation is that you cannot see it directly, which can result in potential exposure when working in those environments. Generally, semiconductor detectors have better energy resolution than scintillation detectors, but the maximum achievable count rates are limited by long pulse signals. Whereas some high pure germanium detectors have been developed to operate at high count rates, and these HPGe detectors could obtain gamma-ray spectra at high count rates exceeding 1 Mcps. However, HPGe detectors require cooling devices to reduce the leak currents, which becomes disadvantageous when developing portable radiation detectors. Furthermore, chemicalcompound semiconductor detectors made of cadmium telluride and cadmium zinc telluride are popular, because they have good energy resolution and are available at room temperature. However, CdTe and CZT detectors develop irradiation-induced defects under intense gamma-ray fields. In this Review, we start with the fundamentals of gamma rays detection and review the recent developments in scintillators gamma-ray detectors. The key factors affecting the detector performance are summarized. We also give an outlook on the field, with emphasis on the challenges to be overcome.