Molten salts have gained significant attention as a potential medium for heat transfer or energy storage and as liquid nuclear fuel, owing to their superior thermal properties. Various fluoride- and chloride-based salts are being explored as potential liquid fuels for several types of molten salt reactors (MSRs). Among these, chloride-based salts have recently received attention in MSR development due to their high solubility in actinides, which has the potential to increase fuel burnup and reduce nuclear water production. Accurate knowledge of the thermal physical properties of molten salts, such as density, viscosity, thermal conductivity, and heat capacity, is critical for the design, licensing, and operation of MSRs. Various experimental techniques have been used to determine the thermal properties of molten salts, and more recently, computational methods such as molecular dynamics simulations have also been utilized to predict these properties. However, information on the thermal physical properties of salts containing actinides is still limited and unreliable. In this study, we analyzed the available thermal physical property database of chloride salts to develop accurate models and simulations that can predict the behavior of molten salts under various operating conditions. Furthermore, we conducted experiments to improve our understanding of the behavior of molten salts. The results of this study are expected to contribute to the development of safer and more efficient MSRs.

Radioactive wastes, including used nuclear fuel and decommissioning wastes, have been treated using molten salts. Electrochemical sensors are one of the options for in-situ process monitoring using molten salts. However, in order to use electrochemical sensors in molten salt, the surface area must be known. This is because the surface area affects the current of the electrode. Previous studies have used a variety of methods to determine the electrode surface area in molten salts. One method of calculating the electrode surface area is to use the reduction current peak difference between electrodes with known length differences. The method is based on the reduction peak and has the benefit of providing long-term in-situ monitoring of surfaces immersed in molten salt. A number of assumptions have been made regarding this method, including that there is no mass transport by migration or convection; the reaction is reversible and limited by diffusion; the chemical activity of the deposit should be unity; and species should follow linear diffusion. For the purpose of overcoming these limitations, a variety of machine learning algorithms were applied to different voltammogram datasets in order to calculate the surface area. Voltammogram datasets were collected from multiarray electrodes, comprising a multiarray holder, two tungsten rods (1 mm diameter) working electrodes, a quasi-reference electrode, and a counter electrode. The multiarray electrode holder was connected to the auto vertical translator, which uses a servo motor, for changing the height of the rod in the molten salts. To make big and diverse data for training machine learning models, various concentrations of corrosion products (Cr, Fe) and fission products (Eu, Sm) in NaCl-MgCl2 eutectic salts were used as electrolyte; electrolyte temperatures were 500, 525, 550, 575, and 600°C. This study will demonstrate the potential of utilizing machine learning based electrochemical in situ monitoring in molten salt processing.

The US NRC developed a program called NRCDose3 to evaluates the environmental impact of radiation around nuclear facilities. The NRCDose3 code is a software suite that integrates the functionality of three individual LADTAP II, GASPAR II, and XOQDOQ Fortran codes that were developed by the NRC in the 1980’s and have been in use by the nuclear industry and the NRC staff for assessments of liquid effluent and gaseous effluent, and meteorological transport and dispersion, respectively. Through the integrated program, it is possible to conduct safety assessment and environmental impact assessment from liquid and gaseous effluent when operating permits are granted. In addition to a more user-friendly graphic user interface (GUI) for inputting data, significant changes have been made to the data management and operation to support expanded capabilities. The basic calculation methods of the LADTAP II, GASPAR II, and XOQDOQ have not been changed with this update to the NRCDose3 code. Several features have been added. The previous program used only ICRP-2 dose conversion factor, but the new program can additionally use dose conversion factor of ICRP-30 and ICRP-72. In the previous program, 4 age groups (infant, child, teen, and adult) were evaluated during dose evaluation, but when ICRP-72 was selected, 6 age groups (infant, 1-year, 5-year, 10-year, 15-year, and adult) could be evaluated. In addition, when selecting ICRP-72, many user-modifiable parameters such as food intake and exposure time were added. It will be referred to E-DOSE60, a program currently under development.

In the pressurized water nuclear reactors (PWRs), the upper and bottom head penetration nozzles, the geometric asymmetry of the welded part increases from the center to the outer part, increasing the possibility of defects. For this reason, it is important to perform early detection and management through analysis of defects occurring in the welded parts of upper and bottom penetration nozzles of reactor vessel. However, it is very difficult to operate boat sampling of the welding area because the spacing of the penetration nozzle of the bottom head of the reactor is very narrow. In addition, it is more difficult to collect welded specimens of bottom penetration nozzles by electrical discharge machining in a boric acid water environment of nuclear reactor. In this work, to overcoming these technical difficulties, we developed a boat sampling robot system, which is composed of the specimen collection electrode head, borate-mediated discharge electrode and control system. Also, we performed basic performance tests and summarize the results.

In the event of a radioactive release, it is essential to quickly detect and locate the source of the release, as well as track the movement of the plume to assess the potential impact on public health and safety. Fixed monitoring posts are limited in their ability to provide a complete picture of the radiation distribution, and the information they provide may not be available in real-time. This is why other types of monitoring systems, such as mobile monitoring, aerial monitoring, and personal dosimeters, are also used in emergency situations to complement the information provided by fixed monitoring posts. Also, the monitoring system can be improved by using the Kriging technique, which is one of the interpolation methods, to predict the radiation dose in the relevant districts. This can be achieved by utilizing both the GPS information and the radiation dose measured at a particular point. The Kriging method involves estimating the value between different measurement points by considering the distance between them. The model used GPS and radiation data that were measured around the Hanbit NPP. The data were collected using a radiation measuring detector on a bus that traveled around the NPP area at 2-second intervals for one day. From the collected data, 200 data points were randomly selected for analysis, excluding the data measured at the bus garage out of a total of 16,550 data points. The average dose of the daily measurement data was 117.94 nSv/h, and the average dose of the 200 randomly extracted data was 119.17 nSv/h. The GPS and radiation dose data were utilized to predict the radiation dose around the Yeonggwang area where the Hanbit NPP is located. In the event of an abnormal release of radioactive material, it can be difficult to accurately determine the dose unless a monitoring measurement point is present. This can delay the rapid evacuation of residents during an emergency situation. By utilizing the Kriging model to make predictions, it is anticipated that more accurate dose predictions can be generated, particularly during accident scenarios. This can aid in the development of appropriate resident protection measures.

In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

To develop technology for extracting energy resources from seawater, we first investigated the research experiences of domestic experts. The survey items included the types of adsorbents that can adsorb dissolved resources in seawater, the subjects of experiments, and the scope of research. We divided the types of adsorbents into organic and inorganic categories and compared their adsorption performance. We also examined how adsorption experiments were conducted using simulated solutions and confirmed whether there were any experiences of conducting experiments in actual seawater. A total of 14 domestic research papers on extracting dissolved resources from seawater were reviewed, excluding topics such as removing dissolved resources from seawater and seawater desalination. This review provides an understanding of domestic research trends and will be helpful in setting directions for future research and development.

Radioactive Oxide is formed on the surface of the coolant pipe of the nuclear power plant. In order to remove the oxide film that is formed on the surfaces of the coolant pipe, chemical and physical decontamination technologies are used. The disadvantage of traditional technologies is that they produce secondary radioactive wastes. Therefore, in this study, the short-pulsed laser eco-friendly technology was used in order to reduce the production of secondary radioactive wastes. It was also used to minimize the damage that was caused to the base material and to remove the contaminated oxide film. The study was carried out using a Stainless steel 304 specimen that was coated with nickel-ferrite particles. Additionally, a transport robot was 3D modeled and manufactured in order to efficiently remove the oxide film from the coolant pipe of the nuclear power plant. The transport robot has a fixed laser head to move inside the horizontal and vertical pipes. The rotating laser head removes the contaminated oxide film on the inner surface of the coolant pipe. In the future, as a condition of the 1064nm short-pulsed laser ablation technique determined by basic analysis, we plan to analyze whether the transport robot is applicable to the radiation contamination site of the nuclear power plant.

Decommissioning plan of nuclear facilities require the radiological characterizations and the establishment of a decommissioning process that can ensure the safety and efficiency of the decommissioning workers. By utilizing the rapidly developed ICT technology, we have developed a technology that can acquire, analyze, and deliver information from the decommissioning work area to ensure the safety of decommissioning workers, optimize the decommissioning process, and actively respond to various decommissioning situations. The established a surveillance system that monitors nuclide inventory and radiation dose distribution at dismantling work area in real time and wireless transmits data for evaluation. Developed an evaluation program based on an evaluation model for optimizing the dismantling process by linking real-time measurement information. We developed a technology that can detect the location of dismantling workers in real time using stereovision cameras and artificial intelligence technology. The developed technology can be used for safety evaluation of dismantling workers and process optimization evaluation by linking the radionuclides inventory and dose distribution in dismantling work space of decommissioning nuclear power plant in the future.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

The concrete structure of a nuclear power plant is a major safety structure that performs shielding functions to block radioactive materials and radiation, heat removal, and isolation functions. Therefore, concrete structures of nuclear power plants must prove structural safety from immediately after construction to dismantling, and a representative method for this is to investigate compressive strength. The compressive strength and specimen standards of concrete structures are specified in ASTM C 42/C 42M, and samples must be obtained through core drilling in order to collect samples according to this standard. However, commercial equipment requires anchor installation work causes radiation dust generation. Even commercial products have developed equipment that does not require anchor installation work, but it can only be applied to flat walls and cannot be applied to curved walls such as bioshields. To solve this problem, a method of fixing to the scaffolding pipe was designed. The equipment developed based on this method fundamentally blocks the generation of radioactive dust. The vertical position can be adjusted using guide shafts and jack screws, and the horizontal position can be adjusted using scaffolding clamps. In addition, the distance between the installation location and the wall can be adjusted by adjusting the scaffolding clamp location of the device. Lastly, it can be rotated to the left and right, so that even on a curved wall, the sampling position can be performed perpendicular to the wall. Core drills that take specimens for measuring compressive strength use the wet type. Core drilling by wet type in radioactively contaminated concrete leads to the disposal of sludge as radioactive waste. Water supplied during core drilling is scattered in all directions by the rotation of the core drill bit, which causes radiation exposure to workers, so measures must be taken to ensure that the water does not splash and gather in one place. Nileplant Co., Ltd. has developed a sludge collection device that can be used with a core drilling device. It can be inserted into a 4-inch core drill bit to meet the specimen regulations of ASTM C 42/C 42M, and nylon resin was used as a material to withstand friction with water, and the wall of the drainage part was thickened to increase durability. Based on these results, it is expected to be able to work more quickly and safely when collecting core drilling samples of radioactively contaminated concrete or radiation and concrete.

Various types of tanks are used in nuclear power plants, and sludge composed of various organic substances and inorganic oxides contaminated with radioactive materials may be present at the bottom of a tank of a radioactive waste treatment device. In addition, glassy and fixative oxide contamination layers are accumulated on the inner wall of the tank depending on the tank material, usage and degree of oxidation. Such contaminated sludge is the main cause of radiation exposure to workers when dismantling nuclear power plant tanks. In addition, the waste filters generated by filtration of contaminated sludge is treated as secondary radioactive waste, and this radioactive waste not only occupies a lot of disposal space, but also the disposal cost is continuously increasing. Therefore, it is necessary to develop a technology that does not generate waste filters as much as possible. To solve this problem, NILEPLANT Co., Ltd. registered a patent named “Filtering apparatus” based on previous research and manufactured a rotary filtration membrane device through detailed design. The rotary filtration membrane device is composed of three or more multiple rotary filtration membranes, and can remove fine particles in wastewater as well as sludge accumulated inside a radioactive contamination tank. In addition, considering the site characteristics of special conditions such as nuclear power plants, it was designed to show excellent performance in removing fine particles while minimizing the area where the device is installed. The rotary filtration membrane device is designed and manufactured as a double cylinder structure that combines a hydro cyclone filter type body and an inner partition wall, and is equipped with a filter cloth-based rotary cylinder filter to process sludge through the filter cloth in addition to inertial. In addition, the patented principle enables self-backwashing without stopping the filtration process, extending the life of the filter and minimizing waste filters. The filtration performance, self-backwashing function, and sludge behavior of the rotary filtration membrane device manufactured based on the detailed design were evaluated through experiments, and improvements to obtain more effective filtration performance were derived. Accordingly, it is expected that the more improved rotary filtration membrane device can be effectively used to remove sludge generated during the dismantling of nuclear power plants in the future.

Laser scabbling has the potential to be a valuable technique capable of effectively decontaminating highly radioactive concrete surface at nuclear decommissioning sites. Laser scabbling tool using an optical fiber has a merits of remote operation at a long range, which provides further safety for workers at nuclear decommissioning sites. Furthermore, there is no reaction force and low secondary waste generation, which reduces waste disposal costs. In this study, an integrated decontamination system with laser scabbling tool was employed to test the removal performance of the concrete surface. The integrated decontamination system consisted of a fiber laser, remote controllable mobile cart, and a debris collector device. The mobile cart controlled the translation speed and position of the optical head coupled with 20 m long process fiber. A 5 kW high-powered laser beam emitted from the optical head impacted the concrete block with dimensions of 300 mm × 300 mm × 80 mm to induce explosive spalling on its surface. The concrete debris generated from the spalling process were collected along the flexible tube connected with collector device. We used a three-dimensional scanner device to measure the removed volume and depth profile.

Laser cutting has been recognized as one of key techniques in dismantling nuclear power plants as it has several advantages such as a remote operation and a reduced secondary waste. However, it generates a significant amount of aerosols that can pose a health risk to workers and further induce environmental pollution during the cutting operation. Thus, understanding the aerosol characteristics generated by the laser cutting is crucial for implementing an effective cutting operation and reducing the exposure to these hazardous particles. In this work, we established a methodology to collect the aerosols and investigate their properties in the laser cutting operation. We built an integrated laser cutting system for aerosol analyses, consisting of a high-power laser cutting module, a metal sample holder, an aerosol collector, and a closed chamber. We expect that this system will offer an opportunity for in-depth understanding of the aerosol properties, by connecting it with desired type of aerosol analysis platforms, and further safe dismantling operation of the nuclear power plants.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

Metakaolin-based geopolymers have shown promise as suitable candidates for 14C immobilization and final disposal. It has been shown that the physicochemical properties of metakaolin wasteforms meet, and often far exceeding, the strict compression strength and leaching acceptance criteria of the South Korea radioactive waste disposal site. However, it is not possible to analyze and characterize the internal structure of the geopolymer wasteform by conventional characterization techniques such as microscopy without destruction of the wasteform; an impractical solution for inspecting wasteforms destined for final disposal. Internal inspection is important for ensuring wastes are homogenously mixed throughout the wasteform and that the wasteform itself does not pose any significant defects that may have formed either during formulation and curing or as a result of testing prior to final disposal. X-ray Computed Tomography (XCT) enables Non-Destructive Evaluation (NDE) of objects, such as final wasteforms, allowing for both their internal and external, characterization without destruction. However, for accurate quantification of an objects dimensions the spatial resolution (length and volume measures) must be know to a high degree of precision and accuracy. This often requires extensive knowledge of the equipment being used, its precise set-up, maintenance and calibration, as well as expert operation to yield the best results. A spatial resolution target consists of manufactured defects of uniformed dimensions and geometries which can be measured to a high degree of accuracy. Implementing the use of a spatial resolution target, the dimensions of which are known and certified independently, would allow for rapid dimensional calibration of XCT systems for the purpose of object metrology. However, for a spatial resolution target to be practical it should be made of the same material as the intended specimen, or at least exhibit comparable X-ray attenuation. In this study, attempts have been made to manufacture spatial resolution targets using geopolymer, silica glass, and alumina rods, as well as 3D printed materials with varying degrees of success. The metakaolin was activated by an alkaline activator KOH to from a geopolymer paste that was moulded into a cylinder (Diameter approx. 25 mm). The solidified geopolymer cylinder as well as both the silica glass rod and alumina rod (Diameter approx. 25 mm) we cut to approximately 4 mm ± 0.5 mm height with additional end caps cut measuring 17.5 mm ± 2.5 mm height. All parts were then polished to a high finish and visually inspected for their suitability as spatial resolution targets.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.