In this review, the regulatory mechanisms of autophagy were described, and its interaction with apoptosis was identified. The role of autophagy in embryogenesis, tooth development, and cell differentiation were also investigated. Autophagy is regulated by various autophagy-related genes and those related to stress response. Highly active autophagy occurrences have been reported during cell differentiation before implantation after fertilization. Autophagy is involved in energy generation and supplies nutrients during early birth, essential to compensate for their deficient supply from the placenta. The contribution of autophagy during tooth development, such as the shape of the crown and root formation, ivory, and homeostasis in cells, was also observed. Genes control autophagy, and studying the role of autophagy in cell differentiation and development was useful for understanding human aging, illness, and health. In the future, the role of specific mechanisms in the development and differentiation of autophagy may increase the understanding of the pathological mechanisms of disease and development processes and is expected to reduce the treatment of various diseases by modulating the autophagic phenomenon.

Currently, non-volatile nuclides such as 94Nb, 99Tc, 90Sr, 55Fe, and 59/63Ni are used a sequential separation. In this study, we developed a separation for 99Tc and 90Sr by a carbonate precipitation. Sodium Carbonate (Na2CO3) was inserted in the aqueous sample from a Dry Active Waste (DAW) and a carbonate precipitation was produced. The precipitate is composed of di- or tri-valent element such as Co, Sr, Fe, Ni and the supernatant is composed of mono-valent element (Cs) and anion materials (ReO4 -, TcO4 -). In DAW, it was confirmed that the recovery of 90Sr (precipitate) and 99Tc (supernatant) were > 90%, respectively. The precipitate and supernatant separated by using a Sr-resin and an anion-exchange resin, respectively. The separated samples were measured by a Liquide Scintillation Counter (LSC, 90Sr) and Induced-Coupled Plasma-Mass Spectroscopy (ICPMS, 99Tc).

I-129 is one of the imporant nuclides that must be determined in the disposal process of radioactive waste in many countries. This radionuclide emits gamma-ray and x-ray photons within the energy range of 29 to 39 keV, consequently, an x-ray detector with high resolution performance is required for the analysis of I-129 activity. An n-type coaxial HPGe detector exhibits higher efficiency characteristics compared to a planar-type HPGe detector, however, its resolution is lower than a planar type. So it is difficult to completely deconvolute and fit the gamma-ray and xray peaks in the spectrum using a general gamma-ray spectrum analysis program such as GammaVision. To address this problem, in a previous study introduced the developed algorithm for the fitting and analysis of I-129 gamma-ray and x-ray spectum by fixing their emission ratios. In this study, we improved the algorithm by considering the variation of the efficiency in the HPGe spectrum, which reflects the actual HPGe crystal condition. And algorithm tests were performed using measured I-129 sample spectra with interfering nuclides acting as background curve are introduced.

One of the important components of a nuclear fuel cycle facility is a hot cell. Hot cells are engineered robust structures and barriers, which are used to handle radioactive materials and to keep workers, public, and the environment safe from radioactive materials. To provide a confinement function for these hot cells, it is necessary to maintain the soundness of the physical structure, but also to maintain the negative pressure inside the hot cell using the operation of the heating, ventilation, and air conditioning (HVAC) systems. The negative pressure inside the hot cells allows air to enter from outside hot cells and limits the leakage of any contaminant or radioactive material within the hot cell to the outside. Thus, the HVAC system is one of the major components for maintaining this negative pressure in the hot cell. However, as the facility ages, all the components of the hot cell HVAC system are also subject to age-related deterioration, which can cause an unexpected failure of some parts. The abnormal operating condition from the failure results in the increase of facility downtime and the decrease in operating efficiency. Although some major parts are considered and constructed in redundancy and diversity aspects, an unexpected failure and abnormal operating condition could result in reduction of public acceptance and reliability to the facility. With the advent of the 4th Industrial Revolution, prognostics and health management (PHM) technology is advancing at a rapid pace. Korea Hydro & Nuclear Power, Siemens, and other companies have already developed technologies to constantly monitor the integrity of power plants and are applying the technology in the form of digital twins for efficiency and safety of their facility operation. The main point of PHM, based on this study, is to monitor changes and variations of soundness and safety of the operation and equipment to analyze current conditions and to ultimately predict the precursors of unexpected failures in advance. Through PHM, it would be possible to establish a maintenance plan before the failure occurs and to perform predictive maintenance rather than corrective maintenance after failures of any component. Therefore, it is of importance to select appropriate diagnostic techniques to monitor and to diagnose the condition of major components using the constant examination and investigation of the PHM technology. In this study, diagnostic techniques are investigated for monitoring of HVAC and discussed for application of PHM into nuclear fuel cycle facilities with hot cells.

As nuclear decommissioning ventures become increasingly complex, the role of digitalization in facilitating and enhancing these operations is becoming indispensable. This transition to a more digitized approach presents a myriad of advantages, including: augmented avenues for data acquisition, analysis, and visualization to bolster dismantling strategies; simulations in virtual environments for operator training; precise forecasting of future waste emergence, culminating in refined cost estimations; and more immersive decommissioning visualizations for both operators and external stakeholders. Salient benefits conferred by the integration of digital technologies in decommissioning encompass improved collaboration, enriched knowledge transfer, clarity regarding present technological constraints, insights into key influencing factors, clearer criteria for technology selection, and a profound understanding of the potential challenges and merits of a broader incorporation of digital tools in decommissioning endeavors. Of paramount importance is the opportunity presented for superior workforce training and safety measures, exemplified by ALARAbased planning. Amidst the myriad facets of digital adoption, 3D modeling of nuclear facilities derived from laser-scanned point clouds stands out as a pivotal domain in the digitalization. The transformation of intricate point cloud data into a comprehensible 3D mesh remains the crux of this paper. The process of mesh generation, despite being simpler than its counterpart of converting to a 3D solid model, is crucial for multiple reasons. The resultant 3D mesh offers an enhanced visual representation compared to a sparse point cloud, paving the way for improved spatial perception. Furthermore, it serves as a rudimentary tool for approximating component volumes and the ensuing waste, thereby playing an instrumental role in waste manipulation strategies, notably in collision detection. This paper delves deep into the nuances of mesh generation, conducting an parametric study of mesh conversion algorithms, including down-sampling rates. Through this rigorous examination, we endeavor to shed light on optimal methodologies, hoping to catalyze advancements in the digitalization of nuclear decommissioning processes.

KEPCO KPS is the contractor for the full system decontamination (FSD) of Kori Unit 1 and under preparation such as modification, lay out for equipment installation, setting up tie-in/out point for chemical injection and way to pressurize the system, of its successful performance. In this research, KPS introduced how KPS has designed and prepared for the FSD project and how will the chemical decontamination process be implemented. As described in the previous research, chemical decontamination process is planned to be conducted for three cycles and each cycle is consisted of oxidation, reduction, decomposition, and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of IX by the heat. If the decontamination result does not meet the target DF and the dose rate, additional cycle can be conducted. Expected volume of process water for FSD is 200 m3. Three systems have been designated as decontamination targets: reactor coolant system (RCS), residual heat removal system (RHRS), chemical volume control system (CVCS). For the steady flow rate, existed plant equipment such as reactor coolant pump (RCP) will be operated and modifications on some components will be conducted. Due to the limited space for installation, decontamination equipment and other resources are distributed to three different places. KPS designed the layout of equipment installed inside the containment vessel. The layout contains the information of shielding for highly radiated equipment such as IX and filter skid.

The thermal treatment of radioactive waste attracts great attention. The thermal treatment offers lots of advantages, such as significant volume reduction, hazard reduction, increase of disposal safety, etc. There are various thermal technologies to waste. The developed technologies are calcination, incineration, melting, molten salt oxidation, plasma, pyrolysis, synroc, vitrification, etc. The off-gas treatment system is widely applied in the technologies to increase the safety and operation efficiency. The thermal treatment generates various by-product and pollutants during the process. The dust or fly ash are generated as a particulate from almost every radioactive waste. The treatment of PVC related components generates hydrogen chloride, which usually brings corrosion of facility. The treatment of rubber and spent resin generates sulfur oxide, SOx. The treatment of nitrile rubber generates nitrogen oxide, NOx. The incomplete combustion of radioactive waste usually generates carbon oxide, COx. The process temperature also affects the generation of off gas, such as NOx and/or COx. Various off gas treatment components are organized for the proper treatment of the previously mentioned materials. In this study systematical review on off gas treatment will be reported. Also, worldwide experiences and developed facility will be reported.

The primary purpose of high temperature process of radioactive waste is to satisfy the waste acceptance criteria and volume reduction. The WAC offers the guideline of waste form fabrication process. The WAC is defined as quantitative or qualitative criteria specified by the regulatory body, or specified by and operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. The main objective of WAC is to protect staff and general public and environment by the containment of radioactive material, limit external radiation level, and prevent criticality. The WAC also offers systematic management of radioactive waste by standardization of waste management operations, facilitation waste tracking, ensure safe and effective operation of operating facilities, etc. Since the high temperature process for radioactive waste is considered in many countries, lots of codes and standards are considered. In many WACs, compressive strength, thermal cycle stability, radiation exposure stability, free liquid, and leachability are evaluation to understand the effect of solidified form to the disposal facility. In this paper, systematical review on waste form will be discussed. In addition, brief result of characterization of waste form will be compared.

Tritium is radioactive isotope, emitting beta ray, released as tritiated water from nuclear power plants. Due to the danger of radioactive isotope, the appropriate separation of tritium is essentially carried out for environment and safety. Further, it is also promising material for energy production and research. The tritiated water can be treated by diverse techniques such as water distillation, cryogenic distillation, Girdler-sulfide process, and catalytic exchange. After treatment, it is more desirable to convert as gas phase for storage, comparing to liquid phase. However, achieving complete separation of hydrogen gases with very similar physical and chemical properties is significantly challenging. Thus, it is necessary to develop materials with effective separation properties in gas separation. In this presentation, we present hydrogen isotope separation in the gas phase using modified mesoporous silica. Mesoporous silica is a form of silica that is characterized by its mesoporous structure possessing pores that range from 2 to 50 nm in diameter. This material can be functionalized to selectively capture and separate molecules having specific size and affinity. Here, the silver and copper incorporated mesoporous silica was synthesized to tailor a chemical affinity quantum sieving effect, thereby providing separation efficiency in D2/H2. The adsorption quantities of H2 and D2 were determined by sorption study, and the textural properties of each mesoporous silica were analyzed using N2 physisorption. The selectivity (D2/H2) in diverse feed composition (1:1, 1:9, and 1:99 of D2/H2) was estimated by applying ideal adsorbed solution theory to predict the loading of the gas mixture on bare, Ag- and Cu-mesoporous silica based on their sorption study. Further, the performance of each mesoporous silica was evaluated in the breakthrough adsorption under 1:1 mixture of D2 and H2 at 77 K.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

The Derived Concentration Guideline Level (DCGL) is required to release the facility from the nuclear safety act at the stage of site restoration of the decommissioning nuclear power plant. In order to evaluate DCGL, there are various requirements, and among them, the selection of input parameters based on the application scenario is the main task. Especially, it is important to select input parameters that reflect site characteristics, and at this time, a single deterministic value or a probabilistic distribution can be applied. If it is inappropriate to apply a particular single value, it may be reasonable to apply various distributions, and the RESRAD code provides for evaluation using probabilistic methods. Therefore, this study aims to analyze the difference between the application of the deterministic method and the application of the probabilistic method to the area and thickness of the contaminated zone among the site characteristics data. This study analyzed the thickness and area of the contaminated zone, and in the case of thickness, the deterministic method was applied by changing the thickness at regular intervals from the minimum depth considered by MARSSIM to the thickness of the unsaturated zone identified in previous research data. In addition, a probabilistic analysis was performed by applying a distribution to the thickness of contaminated zone. Second, for the area of the contaminated zone, the dose was evaluated for each area in consideration of the areas to be considered when deriving Area Factor (AF), and the resulting change in DCGL was observed. As a result, the DCGL tends to decrease as the thickness increases, and it seems to be saturated when the thickness exceeds a certain thickness. Therefore, It was confirmed that the level of saturated values is similar to that of entering a probabilistic distribution, and in the case of a parameter that is reasonable to enter as a distribution rather than as a single value, it is sufficiently conservative to perform a probabilistic evaluation. In the case of area change, the DCGL evaluation result showed that the DCGL increased as the scale decreased. The magnitude of the change varies depending on the characteristics of each radionuclide, and in the case of radionuclides where external exposure gamma rays have a major exposure effect, the change is relatively small. It can be seen that the change in DCGL according to the area has the same tendency as the AF applicable to the survey unit for small survey units applied in the final status survey.

As unit 1 of Kori was permanently shut down in June 2017, domestic nuclear industry has entered the path of decommissioning. The most important thing in decommissioning is cost reduction. And volume reduction of radioactive waste is especially important. According to the IAEA report, more than 4,000 tons of metallic waste is generated during the decommissioning of a 1,000 MWe reactor and most of these wastes are LLW or VLLW. To reduce amount of metallic waste dramatically, we should choose efficient decontamination method. In this study, we conducted dry ice and bead blasting decontamination. We prepared Inconel-600 and STS-304 specimen with dimensions of 30 mm × 30 mm × 5 mm. Loose and fixed contamination was applied on the surface of specimen using SIMCON method. Bead and dry-ice blasting was conducted by spraying alumina and dry ice pellet at the same pressure and distance for the same time. The removal of loose contamination was observed using microscope. It was found that contaminants are significantly removed using both dry ice blasting and bead blasting. However, some abrasive material remained on the surface of specimen. The removal of fixed contamination was verified by weight comparison before and after experiment and cobalt concentration comparison before and after experiment using X-ray Fluorescence Spectroscope (XRF). At least 90% of the cobalt was removed, but some abrasive particle was also remained on the surface of specimen. In this study, it is confirmed that the effectiveness of manufacturing a large-scale abrasive decontamination facility, and it is expected that this technology can be used to effectively reduce the amount of metallic waste generated during decommissioning.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

As the acceptance criteria for low-intermediate-level radioactive waste cave disposal facilities of Korea Radioactive Waste Agency (KORAD) were revised, the requirements for characterization of whether radioactive waste contains hazardous substances have been strengthened. In addition, As the recent the Nuclear Safety and Security Commission Notice (Regulations on Delivery of Low- Medium-Level Radioactive Waste) scheduled to be revised, the management targets and standards for hazardous substances are scheduled to be specified and detailed. Accordingly, the Korea Atomic Energy Research Institute (KAERI) needs to prepare management methods and procedures for hazardous substances. In particular, in order to characterize the chemical requirements (explosiveness, ignitability, flammability, corrosiveness, and toxicity) contained in radioactive waste, it must be proven through documents or data that each item does not contain hazardous substances, and quality assurance for the overall process must be provided. In order to identify the characteristics of radioactive waste that will continue to be generated in the future, KAERI needs to introduce a management system for hazardous substances in radioactive waste and establish a quality assurance system. Currently, KAERI is thoroughly managing chelates (EDTA, NTA, etc.), but the detailed management procedures for hazardous substances related to chemical requirements in radioactive waste in the radiation management area specified above are insufficient. The KAERI’s Laboratory Safety Information Network has a total periodic regulatory review system in place for the purchase, movement, and disposal of chemical substances for each facility. However, there is no documents or data to prove that the hazardous substances held in the facility are not included in the radioactive waste, and there are no procedures for managing hazardous substances. Therefore, it is necessary to establish procedures for the management of hazardous substances, and we plan to prepare management procedures for hazardous substances so that chemical substances can be managed according to the procedures at each facility during preliminary inspection before receiving radioactive waste. The procedure provides definitions of terms and types of management targets for each characteristic of the chemical requirements specified above (explosiveness, ignition, flammability, corrosiveness, and toxicity). In addition, procedure also contains treatment methods of radioactive waste generated by using hazardous substances and management methods of in/out, quantity, history of that substances, etc. As the law is revised in the future, management will be carried out according to the relevant procedures. In this study, we aim to present the hazardous substance management procedures being established to determine whether radioactive waste contains hazardous substances in accordance with the revised the notice and strengthened acceptance criteria. Through this, we hope to contribute to improving reliability so that radioactive waste could be disposed of thoroughly and safely.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR 1&2), the first research reactors in South Korea, began in 1997 and the decommissioning status is currently proceeding with phase 3. It is expected that more than 5,000 tons of dismantled wastes will be generated as the contaminated building is demolished. Since these dismantled wastes must be disposed of in an efficient method considering economic feasibility, it is desirable to clearance extremely low-level wastes whose contamination is so minimal that the radiological risk is negligible. In Korea, in order to approve the clearance of radioactive waste, it must be proven that the nuclide concentration standards are met or that the dose to individuals and collectives is below the allowable dose value. At the KRR 1&2 decommissioning site, dismantled wastes have been steadily being disposed of through clearance procedure since 2021. Clearance was approved by the Korean Institute of Nuclear Safety (KINS) for one case of concrete waste in 2021 and two cases of metal waste in 2022. In 2023, the clearance of metal waste and asbestos waste has been approved so far, and in particular, this is the first case in Korea for asbestos waste. In this study, we compared the dose assessment methods and results of clearance wastes at the KRR 1&2 decommissioning site from 2021 to present. Dose assessment was conducted by applying the landfill scenario for concrete and asbestos and the recycling scenario for metal waste. The calculation codes used were RESRAD-onsite 7.2 and RESRAD-recycle 3.10. The dose conversion factors (DCF) for each age group (infant, 1y, 5y, 10y, 15y, adult) of the target nuclide used the values presented in ICRP-72, and in particular, geo-hydrological data of the actual landfill site was used as an input factor when evaluating landfill scenarios. As a result of the dose assessment, when landfilling concrete wastes in 2020, the personal dose and collective dose were evaluated the most at 2.80E+00 μSv/y and 4.83E-02 man·Sv/y, respectively.

At domestic nuclear power plant, concrete containers are stored to store waste generated before waste acceptance criteria (WAC) was established. Concrete container store concentrated waste liquid and waste resin. In order to disposal radioactive waste to a disposal site, it is necessary to conduct a characteristic evaluation inside the waste to check whether it satisfies the WAC. Two types of concrete containers are stored: round and square. The round type is filled with one 200-liter drum, and the square type is filled with four 200-liter drums. In the case of a round shape, the top lid is fastened with bolts, so it is possible to collect samples after opening the top lid without the need for additional equipment. However, in the case of a square shape, there is no top lid, and concrete is poured to cure the lid, so the separate equipment for characteristic evaluation is required. It is necessary to install a workstation for sample collection on the top of the concrete container, equipment for coring the top of the concrete container, and a device to prevent concrete dust scattering. Currently, the design of equipment for evaluating the characteristics of concrete containers has been completed, and equipment optimization through mock-up test will be performed in the future.

In nuclear power plant environments, the analysis of gamma-emitting waste materials with complex shapes can be challenging. ISOCS (In-Situ Objective Counting System) is employed to measure the gamma-emitting radionuclide concentrations. However, it is crucial to validate the accuracy of ISOCS measurements. This study aims to validate the accuracy of ISOCS measurement results for spent filters. The ISOCS measurement process begins with modeling and efficiency calculations of the target spent filters using ISOCS software. ISOCS offers the advantage of direct measurement assessment by incorporating shielding materials and collimators into the detector efficiency calculation during the modeling process, without the need for separate efficiency correction sources. To validate the accuracy of ISOCS measurement results, the measured radioactivity values were used as input data for the MicroShield computer code to derive dose rates. These dose rates were then compared to the dose rates measured on-site, confirming the reliability of ISOCS measurements. In the field, ISOCS gamma measurements and surface dose rates were measured for three Cavity filters and four RCP Seal Injection filters. The measured dose rate for the Cavity filters was around 270 Svhr, and the computed values using MicroShield showed an error of approximately 12%. Despite modeling and calculation errors in computer analysis and potential uncertainties in the measurement environment and instrument, the computed values closely matched the measured values. However, the measured dose rate for the RCP Seal Injection filters ranged 2.9~8 Svhr, which is very low and close to background levels. When compared to the results of computer analysis, an error ranging from 27% to 97% was observed. It is concluded that validating the accuracy in the low dose rate range close to background levels is challenging through a comparison of calculated and measured dose rates.