Hydrogen-bonded organic frameworks (HOFs) are a new type of porous crystalline material that are constructed by intermolecular hydrogen bonding of organic building blocks to form twodimensional (2D) and three-dimensional (3D) crystalline networks. High-quality HOF single crystals are easily grown for direct superstructure analysis using single crystal X-ray diffraction, which is essential for revealing the relationship between structure and properties. The unique advantages of HOF, such as high crystallinity, porosity and fast regeneration, have allowed it to be used in a variety of applications including catalysis and gas separation. Squaric acid (SQA) is a non-carboxylic, organic acid with proton donor and acceptor ability which is known to take on a variety of coordination modes with metal ions. Pyrazine is a six-membered aromatic heterocycle bearing two nitrogen atoms, which has sp2 hybridized C atoms with C-H hydrogen bonds. This work describes the synthesis and structural characteristics of HOF based on squaric acid and pyrazine. Based on single crystal X-ray diffraction data, this MOF crystallizes in the triclinic P-1 space group. Each asymmetric unit is composed of H2SQ and pyrazine. All squaric acid molecules share one H atom with the N atom of pyrazine molecules. The layer distance between nearby O atoms from squaric acid in different layers equals 5.29 Å. Also, our HOF showed high adsorption capacity the during experiments. The composition and comparative characteristics of HOF are given using SCXRD, PXRD, SEM and UV-vis.

The sorption/adsorption behavior of radionuclides, usually occurring at the solid-water interface, is considered to be one of the primary reactions that can hinder the migration of radiotoxic elements contained in the spent nuclear fuel. In general, various physicochemical properties such as surface area, cation exchange capacity, type of radionuclides, solid-to-liquid ratio, aqueous concentration, etc. are known to provide a significant influence on the sorption/adsorption characteristics of target radionuclides onto the mineral surfaces. Therefore, the distribution coefficient, Kd, inherently shows a conditiondependent behavior according to those highly complicated chemical reactions at the solid-water interfaces. Even though a comprehensive understanding of the sorption behavior of radionuclides is significantly required for reliable safety assessment modeling, the number of the chemical thermodynamic model that can precisely predict the sorption/adsorption behavior of radionuclides is very limited. The machine-learning based approaches such as random forest, artificial neural networks, etc. provide an alternative way to understand and estimate complicated chemical reactions under arbitrarily given conditions. In this respect, the objective of this study is to predict the sorption characteristics of various radionuclides onto major bentonite minerals, as backfill materials for the HLW repository, in terms of the distribution coefficient by using a machine-learning based computational approach. As a background dataset, the sorption database previously established by the JAEA was employed for random forest machine learning calculation. Moreover, the hyperparameters such as the number of decision trees, the number of variables to divide each node, and random seed numbers were controlled to assess the coefficient of determination, R2, and the final calculation result. The result obtained in this study indicates that the distribution coefficients of various radionuclides onto bentonite minerals can be reliably predicted by using the machine learning model and sorption database.

Radioactive contamination of soil on the site of a nuclear facility has a characteristic that radioactive nuclides are adsorbed into the pores between soil particles, making it quite difficult to decontaminate. For this reason, research on the development of various decontamination processes is being actively conducted. In this study, among various decontamination studies, a soil decontamination process using supercritical carbon dioxide was presented. The decontamination process uses supercritical carbon dioxide as the main solvent, which has a higher penetration power than other materials. Therefore, the process consists of the process of desorbing and extracting the target radionuclides between particles of soil. However, since nuclides exist as ions in the soil, polar chelating ligand material was introduced as an additive to nonpolar supercritical carbon dioxide for smooth chemical reactions in the soil. Thereafter, from the viewpoint of improving process continuity and efficiency, an alcohol material was introduced as an auxiliary solvent for liquefaction of chelating ligand in a solid state. Through prior research on the selection of a solvent for liquefaction of chelating ligand, ethanol and 2-propanol were finally selected based on whether the chelating ligand was dissolved. However, if the auxiliary solvent in which the chelating ligand is dissolved is to be combined with radionuclides in the soil, it must first be well dissolved in supercritical carbon dioxide, the main solvent. Therefore, in this study, the solubility of ethanol and 2-propanol in supercritical carbon dioxide was measured and the suitability was evaluated. The temperature conditions were carried out at 40°C, the same as the previously designed decontamination process, and the measurement was conducted by adjusting the pressure and volume through a syringe pump and a variable volume device. In addition, solubility was measured based on the observation of the ‘cloud point’ in which the image becomes cloudy and then bright. As a result of the experiment, several solubility points were measured at a pressure of 150 bar or less. If the flow rate ratio of supercritical carbon dioxide and auxiliary solvent derived from the results is applied to the soil decontamination process, it is expected that the process efficiency will increase in the future.

Radiological characterization is important in decommissioning and dismantling of nuclear facilities, in order to assess the radioactivity concentration, classify the wastes, and secure workers’ safety. The Some components such as Reactor Pressure Vessel (RPV) in nuclear facilities has dose rate higher than Sv/hr, thus in-situ gamma spectroscopy systems suffer from a very high count rate which causes energy resolution degradation, photo-peak shift, and count loss by pile-up and dead-time. The system must be operated in a very high count rate, in order to measure spectra precisely and to quantify radionuclide contents. In order to apply in-situ measurement in high radiation dose rate environment, the sensor, front-end electronics, and data acquisition (DAQ) should be carefully selected and designed as well as precise design of collimators and radiation shield. In this paper, the components of the detector system were selected and performance was evaluated in a high count rate before design the collimator and shield. A LaBr3 coupled with a PMT having short decay time constant (16 nsec) was selected for high count rate application, and two different amplifiers (a conventional charge sensitive preamplifier with 50 usec decay time constant, and wide-band voltage amplifier) were tested. As DAQs, DT5781 (14 bit, 100 MS/s, CAEN) of Pulse Height Analysis (PHA) which is conventionally used signal processing method in the gamma spectroscopy, and DT5730 (14 bit, 500MS/s, CAEN) of Pulse Shape Discrimination (PSD) which is similar to Charge to Digital Convertor (QDC) were used. The number of photons incident to the detector was varied by changing the detector-source distance with Certificate Radiation Material (CRM), and compared to the output count rate. The count rate capability, and energy resolution with different amplifier and DAQ was evaluated. Additionally, the performance of DAQs in extremely high count rate was evaluated with signal data generated by the emulator which can simulate the detector signal waveforms fed into the DAQ based on the measured spectrum.

A sequential column experiment was conducted for uranium removal of excessively high or highly U-contaminated soils, simultaneously. Two pilot-scale acryl columns with a 24 cm ID and 48 cm length were uniformly packed with each U-contaminated soil (both < 2 mm, 119, and 22.4 Bq/g as initial U-238 activities). A column packed with soil contained very high U constant located first then sequentially located second columns with relatively lower U-contaminated soil. Thus the effluents which passed very high U-contaminated soil and having extremely high dissolved U concentration was directly inflowed the second columns. Both columns initially and respectively flushed with demi water (or condensing water of air conditioner generated from radiation controlled area) to saturate and displace the air from the pore space. Elution was carried out with alkaline and acidic solutions, respectively, and sequentially. The uranium removal efficiencies were found and a comparison was made with the pilot soil flushing experiments. During this study, a new approach to reducing acidic flushing waste which is considered the biggest defect of soil washing/flushing was established, and optimal factors were calculated to demonstrate industrial-scale uranium decontamination of soil with high uranium content.

Laser cutting has been attracting attention as a next-generation tool in application for nuclear decommissioning. It enables high-speed cutting of thick metal objects, and its narrow kerf width greatly reduces the amount of secondary waste compared to other cutting methods. In addition, it only requires the relatively small cutting head without any complicated equipment, and long-distance cutting apart from a laser generator is possible using beam delivery through optical fiber. And there is almost no reaction force because it is non-contact thermal cutting. For these reasons, the laser cutting is very advantageous for remote cutting. In laser cutting, the irradiated laser power is absorbed and consumed to melt the material of the cutting target. When the applied laser power is greater than the power consumed for melting, the residual power is transmitted to the back of the cut object. This residual power may unintentionally cut or damage undesired objects located behind the cutting target. In order to prevent this, it is necessary to adjust the laser power for each thickness of the target object to be cut, or to increase the distance between the cut target and the surrounding structures so that the transmitted power density can be sufficiently lowered. In this work, safety study on residual power that penetrates laser-cut objects was conducted. Experimental studies were performed to find safe conditions for irradiation power density that does not cause surface damage to the stainless steel by adjusting the laser power and stand-off distance from the target.

In addition to Korea, various countries such as the United States, the United Kingdom, France, and China are designing small module-type reactors. In particular, a small modular reactor is the power of 300 MWe or less, in which the main equipment constituting the nuclear reactor is integrated into a single container. Depending on the purpose, small modular reactors are being developed to help daily life such as power, heating supply, and seawater desalination, or for power supply such as icebreakers, nuclear submarines, and spacecraft propellants. Small modular reactors are classified according to form. It can be classified into light-water reactors/ pressurized light-water reactors based on technology proven in commercial reactors, and non-lightwater reactors based on fuel and coolant type such as Sodium-cooled Fast Reactor, High temperature gas-cooled reactor, Very high temperature reactor and Moltenn salt reactor. SMRs, which are designed for various purposes, have the biggest difference from commercial nuclear reactors. The size of SMRs is as small as 1/5 of that of the commercial reactors. Several modules may be installed to generate the same power as commercial reactors. Because of the individually operation for each module, load follow is possible. Also, The reactor can be cooled by natural convection because the size is small enough. It is manufactured as a module, the construction period can be reduced. Depending on the characteristics of these SMRs, application for safeguards is considered. There are many things to consider in terms of safeguards. Therefore, it is IAEA inspection or other approaches for SMRs installed and remotely operated in isolated areas, data integrity for remote monitoring equipment to prevent the diversion of nuclear materials, verification method and material accountancy and control for new fuel types and reactors. Since SMR is more compact and technical intensive, safeguards should be considered at the design stage so that safeguards can be efficiently and effectively implemented, which is called the Safeguards by design (SBD) in the IAEA. In this paper, according to the characteristics of SMR, we will analyze the advantages/disadvantages from the point of view of safeguards and explain what should be considered.

Kori unit 1, Korea’s first light-water nuclear power plant, was permanently shut down in June 2017. The operator, Korea Hydro & Nuclear Power Co. (KHNP), submitted a final dismantling plan for Kori unit 1 to the Nuclear Safety and Security Commission (NSSC) in May 2021. Pursuant to this procedure, the NSSC is preparing regulations for the decommissioning stage of large nuclear facilities for the first time in the Republic of Korea. The Korea Institute of Nuclear Non-proliferation and Control (KINAC) is also considering applying regulations on safeguards. Moreover, the International Atomic Energy Agency (IAEA) developed the “International Safeguards Guidelines for Nuclear Facilities under Decommissioning” in 2021. The guidelines describe the detailed application of safeguards measures to be considered when decommissioning nuclear facilities, dismantling essential equipment, and providing relevant information to the IAEA, as well as the scope of IAEA inspections. In addition, Dr. R. Bari of the Brookhaven National Laboratory (BNL) proposed the Facility Safeguardability Assessment (FSA), a methodology that reflects facility characteristics from the design stage to ensure that designers, national regulators, and the IAEA communicate smoothly regarding safeguards measures. The FSA process derives expected problems with safeguards measures considering new nuclear facilities by analyzing the gap of safeguards measures applied to existing similar nuclear facilities. This study uses the existing FSA methodology to predict problems related to safeguards measures when decommissioning nuclear facilities and to analyze deviations from safeguards measure requirements according to IAEA guidelines. To this end, the reference facility is set as an operating pressurized light water reactor; the issues with the safeguards measures are summarized using the FSA Process; and a draft safeguards concept for nuclear facilities under decommissioning is designed. Furthermore, validity is confirmed through a simple analysis of the diversion path, and implications and lessons are derived. Through this, it is possible to anticipate new safeguards measures to be applied when decommissioning nuclear facilities in the Republic of Korea and review problems and considerations in advance.