In this study, we have fabricated the phenolic resin (PR)/polyacrylonitrile (PAN) blend-derived core-sheath nanostructured carbon nanofibers (CNFs) via one-pot solution electrospinning. The obtained core-sheath nanostructured carbon nanofibers were further treated by mixed salt activation process to develop the activated porous CNFs (CNF-A). Compared to pure PAN-based CNFs, the activated PR/PAN blend with PR 20% (CNF28-A)-derived core-sheath nanostructured CNFs showed enhanced specific capacitance of ~ 223 F g− 1 under a three-electrode configuration. Besides, the assembled symmetric CNF28-A//CNF28-A device possessed a specific capacitance of 76.7 F g− 1 at a current density of 1 A g− 1 and exhibited good stability of 111% after 5,000 galvanostatic charge/discharge (GCD) cycles, which verifies the outstanding long-term cycle stability of the device. Moreover, the fabricated supercapacitor device delivered an energy density of 8.63 Wh kg− 1 at a power density of 450 W kg− 1.

Nanofillers, by virtue of their minute size, when incorporated inside a matrix, have the capability to enhance the physical parameters of the complex matrix. Graphene, the wonder material of the twenty-first century, has established itself in the field of nanofillers. However, it still has yet to find its way into the mining industry. This review paper focuses on a novel way of attaining sustainability in mining methodology using graphene as a nanofiller. The implementations can be subdivided into three categories based on their impact—economic, environment, and safety. To achieve economic welfare in mine methodology; Graphene is used to enhance the productivity of machinery. Electric-Heavy Earth Moving Machinery using LiFePO4/Graphene hybrid cathode battery is not only an ideal replacement to fossil-powered vehicles considering the contribution of environmental strain but also a more-efficient model than Electric-Heavy Earth Moving Machinery using conventional Lithium Ion Phosphate Battery batteries. Heavy Earth Moving Machinery having tires of Styrene-Butadiene Rubber/Graphene composite would have better efficacy and longer life cycle than the conventional ones. Graphene derivative Magnetic Graphene Oxide is used to achieve environmental welfare by its implication as an additive in the effluent treatment plant for its capability of removing heavy metal ions and negative-strain bacteria from the mine water. To improve the safety standards of the mine workers, graphene and its derivatives have environmental implications to constitute a safer surrounding concerning precarious situations due to the unpredictable behavior of geomaterials. Graphene can assist in constituting a more economical and reliable slope model as incorporating graphene induces restructuration and improvement in strength parameters. This enables a miner to extract more minerals in tranquility from the resources as there is an increase in compaction and shear strength. A combination of a graphene sheet and auxetic graphene foam can be placed over the blast holes to not only restrict the trajectory of the fly rocks but also attenuate some part of the explosive energy. The objective of this coagulation is to upgrade the traditional practice by replacing the conventional products, and the effect is observed in the form of achieving sustainability in the mine.

Given the limited terrestrial reserves of uranium (approximately 4.6 million tons), exploring alternative resources is necessary to secure a sustainable, long-term supply of nuclear energy. Uranium extraction from seawater (UES) is a potential solution since the amount of uranium dissolved in seawater (approximately 4.5 billion tons) is about 1,000 times that of terrestrial reserves. However, due to the ultra-low concentration of uranium in seawater (approximately 3.3 ppb), making UES economically viable is a challenging task. In this paper, we explore the potential of using thermal discharge from domestic nuclear power plants for uranium extraction. The motivation for this comes from previous research showing that the adsorption capacity of amidoxime-based adsorbents is proportional to the temperature of the seawater in which they are deployed. Specifically, a study conducted in Japan found that a 10°C increase in seawater temperature resulted in a 1.5-fold increase in adsorption capacity.

To analyze the radioactivity of 3H and 14C in miscellaneous radioactive wastes generated from nuclear power plants, a wet digestion method using sulfuric acid is currently used. However, sulfuric acid is classified as a special management material, and there is no disposal method for contaminated radioactive waste. Therefore, research on a thermal decomposition method that can analyze the DAW radioactive waste samples without using sulfuric acid is necessary. In this study, we will cover the final sample amount, sample injection method, and prevention of organic ignition to meet the minimum detection limit requirements of the analysis equipment. Through this research, optimal conditions for the thermal decomposition method for analyzing the radioactivity of 3H and 14C in DAW radioactive wastes generated from nuclear power plants can be derived.

For the final disposal of radioactive waste, concentration of gamma nuclides such as Co-58, Co-60, Cs-137, Nb-94 have to be determined to meet nuclear regulatory requirements. In general, gamma nuclide analysis can be performed with simple sample pretreatment without complicated chemical separation processes due to the characteristics of the nuclide and high resolution of the measuring equipment. However, when the concentration of Co-60 is high in a specific radioactive waste generated at the NPP, the background is increased by the compton continuum of Co-60. That makes it difficult to evaluate accurately Nb-94, which is in the lower energy band than the gamma ray energy region of Co-60 and especially Cs-137, which is used as a key nuclide of scaling factor. In this study, We consider the problem of MDA dissatisfaction or overestimation due to the increased background by Co-60.

When self-disposing of radioactive waste, it is important to follow the acceptable concentration standards for each nuclide set by the Nuclear Safety and Security Commission (NSSC). Gamma-emitting nuclides can be easily analyzed with a simple pretreatment process, but beta-emitting nuclides require a chemical separation procedure to be analyzed for radiochemistry analysis. When analyzing betaemitting nuclides for the purpose of self-disposal, there may be difficulties in radiation detection after the chemical separation process. This is because the concentration of beta nuclides in the sample may be low and some of them may be lost during the chemical separation. Therefore, measurement method of gross-beta activity can be used instead of that of each nuclide to access the compliance of selfdisposal criteria. While a proportional counter is commonly used to measure gross-beta activity, liquid scintillation counting can also be used to measure gross-beta, and we plan to compare the results of both methods.

Many radionuclides emit two or more gamma rays in a cascade once they decay. At this time, gamma rays are detected at the same time, and the signals are overlapped and measured as one added signal. This is called the summing coincidence effect, and it causes an error of more than 10% depending on the detection efficiency, measurement conditions, and target nuclide. It is known to be greater as the efficiency of the detector increases and as the distance between the source and the detector decreases. It is necessary to consider the summing coincidence effect since the efficiency of the HPGe detector owned by the KHNP CRI is as high as 65%. In this study, We would like to propose an appropriate gamma nuclide analysis method for radioactive waste generated from NPP by evaluating the influence on the summing coincidence effect.

To analyze the activity concentration of radionuclides in radioactive sludge samples generated from low- and intermediate-low-level radioactive waste from domestic nuclear power plant, a pretreatment process that dissolves and homogenizes the sample is essential. However, this pretreatment process requires the use of hydrofluoric acid, which makes analysis difficult and challenges users to handle harmful chemicals. Therefore, we aim to minimize the use of hydrofluoric acid by measuring gamma nuclides in the sludge sample without pretreatment process and compare the differences of measurement results according to the sample matrix with and without pretreatment process. We will collect about 0.1 g of the sludge sample, and dissolve it using an acid treatment process after using microwave decomposition. We will then use gamma spectroscopy to check the concentration of nuclides present in the sludge before and after dissolution and consider the effect of the sample matrix.

Confirmation of the thermal behavior of spent fuel is one of the important points in the management of high-level radioactive waste. This is because various fission products exist in spent nuclear fuel, and a management plan according to their behavior is required. Among the fission products, epsilon particles exist in the form of metal deposits and have a great influence on their physical and chemical properties. However, observing the thermal behavior of epsilon particles is important for understanding spent fuel behavior in thermally environment, but it is difficult to maintain a consistent thermal environment. In this work, we report the thermal behaviors study of uranium oxide with epsilon particle using in situ high temperature X-ray diffraction. We measured the variation of temperature on the size of crystalline, which is a cell parameter in the reaction process. And then, the change of lattice parameters is calculated by Rietveld refinement.

Zirconium(Zr) alloys are commonly used in the nuclear industry for applications such as fuel cladding and pressure tubes. To minimize the levels and volumes of radioactive waste, molten salts have been employed for decontaminating Zr alloys. Recently, a two-step Zr metal recovery process, combining electrolysis and thermal decomposition, has been proposed. In the electrolysis process, potentiostatic electrorefining is utilized to control the chemical form of electrodeposits(ZrCl). Although Zr metals are expected to dissolve into molten salts, reductive alloy elements can also be co-dissolved and deposited on the cathode. Therefore, a better understanding of the anodic side’s response during potentiostatic electrorefining is necessary to ensure the purity of recovered Zr and long-term process operation. As the first step, potentiodynamic polarization curves were obtained using Zr, Nb, and Zr-Nb alloy to investigate the anodic dissolution behavior in the molten salts. Nb, which has a redox potential close to Zr, and Zr exhibit active or passivation dissolution mechanisms depending on the potential range. It was confirmed that Zr-Nb alloy also has a passivation region between -0.223 to -0.092 V influenced by the major elements Zr and Nb. Secondly, active dissolution of Zr-Nb was performed in the range of -0.9 to -0.6 V. The dissolution mechanism can be explained by percolation theory, which is consistent with the observed microstructure of the alloy. Thirdly, passivation dissolution of Zr, Nb, and Zr-Nb alloy was investigated to identify the pure passivation products and additional products in the Zr-Nb alloy case. K2ZrCl6 and K3NbCl6 were identified as the pure passivation products of the major elements. In the Zr-Nb alloy case, additional products, such as Nb and NbZr, produced by the redox reaction of nanoparticles in the high viscous salt layer near the anode, were also confirmed. The anodic dissolution mechanism of Zr-Nb alloy can be summarized as follows. During active dissolution, only Zr metal dissolves into molten salts by percolation. Above the solubility near the anode, passivation products begin to form. The anode potential increases due to the disturbance of passivation products on ion flow, leading to co-dissolution of Nb. When the concentration of Nb ion exceeds the solubility, a passivation product of Nb also forms. In this scenario, a high viscous salt layer is formed, which traps nanoparticles of Zr metal, resulting in redox behavior between Zr metal and Nb ion. Some nanoparticles of Zr and Nb metal are also present in the form of NbZr.

A low- and intermediate-level radioactive waste repository contains different types of radionuclides and organic complexing agents. Their chemical interaction in the repository can result in the formation of radionuclide-ligand complexes, leading to their high transport behaviors in the engineered and natural rock barriers. Furthermore, the release of radionuclides from the repository can pose a significant risk to both human health and the environment. This study explores the impact of different experimental conditions on the transport behaviors of 99Tc, 137Cs, and 238U through three types of barrier samples: concrete, sedimentary rock, and granite. To assess the transport behavior of the samples, the geochemical characteristics were determined using X-ray diffraction (XRD), X-ray fluorescence (XRF), Fouriertransform infrared spectroscopy (FTIR), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS), and Brunauer-Emmett-Teller (BET) analysis. The adsorption distribution coefficient (Kd) was used as an indicator of transport behavior, and it was determined in batch systems under different conditions such as solution pH (ranging from 7 to 13), temperature (ranging from 10 to 40°C), and with the presence of organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA). A support vector machine (SVM) was used to develop a prediction model for the Kd values. It was found that, regardless of the experimental parameters, 99Tc may migrate easily due to its anionic property. Conversely, 137Cs showed low transport behaviors under all tested conditions. The transport behaviors of 238U were impacted by the order of EDTA > NTA> ISA, particularly with the concrete sample. The SVM models can also be used to predict the Kd values of the radionuclides in the event of an accidental release from the repository.

The organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA) can enhance the radionuclides’ solubility and have the potential to induce the acceleration of radionuclides’ mobility to a far-field from the radioactive waste repository. Hence, it is essential to evaluate the effect of organic complexing agents on radionuclide solubility through experimental analysis under similar conditions to those at the radioactive waste disposal site. In this study, five radionuclides (cesium, cobalt, strontium, iodine, and uranium) and three organic complexing agents (EDTA, NTA, and ISA) were selected as model substances. To simulate environmental conditions, the groundwater was collected near the repository and applied for solubility experiments. The solubility experiments were carried out under various ranges of pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and concentrations of organic complexing agents (0, 10-5, 10-4, 10-3, and 10-2 M). Experimental results showed that the presence of organic complexing agents significantly increased the solubility of the radionuclides. Cobalt and strontium had high solubility enhancement factors, even at low concentrations of organic complexing agents. We also developed a support vector machine (SVM) model using some of the experimental data and validated it using the rest of the solubility data. The root mean square error (RMSE) in the training and validation sets was 0.012 and 0.016, respectively. The SVM model allowed us to estimate the solubility value under untested conditions (e.g., pH 12, temperature 30°C, ISA 5×10-4 M). Therefore, our experimental solubility data and the SVM model can be used to predict radionuclide solubility and solubility enhancement by organic complexing agents under various conditions.

The operation of nuclear power plants, nuclear waste depositories, and the decontamination and decommissioning of nuclear power plants all have the possibility of generating various kinds of radionuclides that can be formed as gaseous or liquid phases. Among the radionuclides, strontium is considered as most harmful substance due to its abundance in nuclear accident effluent, long half-life, high fission yield, high water solubility, and high mobility in aquatic environment. To remove strontium from aquatic environment, adsorption technique is mainly used with high economic feasibility, efficiency, and selectivity. Previously, we synthesized sodium titanates with mid-temperature hydrothermal method as selective strontium adsorbent in aqueous solution. Moreover, it was demonstrated that synthesized sodium titanates show high strontium adsorption rate with high selectivity with high surface area, pore diameter and volume. Herein, we investigated the surface structure of synthesized sodium titanates before and after strontium adsorption in aqueous solution using scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), and X-ray photoelectron spectroscopy (XPS) analysis. According to SEM and EDS experimental results, aquatic strontium can be adsorbed as surface precipitation with formation of cube-shaped structure, which is quite similar strontium titanate structure crystals onto the surface of sodium titanates. In addition, XPS experimental results revealed that the titanium ions on the surface of sodium titanates were oxidized during strontium surface precipitation process, and the sodium ion on the surface of sodium titanates were exchanged with aquatic strontium ions via ion exchange process during strontium adsorption process.

In response to a regulatory mandate, all nuclear licensees are obligated to establish an information system that can provide essential information in the event of a radiation emergency by connecting the monitoring data of the Safety Parameter Display System (SPDS) or equivalent system to the Korea Institute of Nuclear Safety (KINS). Responding to this responsibility, the Korea Atomic Energy Research Institute (KAERI) has established the Safety Information Transmission System (SITS), which enables the collection and real-time monitoring of safety information. The KAERI monitors and collects safety information, which includes data from the HANARO Operation Work Station (OWS) and the HANARO & HANARO Fuel Fabrication Plant (HFFP) Radioactivity Monitoring System (RMS), and the Environmental Radiation Monitoring System (ERMS) & meteorological data. Currently, the transmission of this safety information to the AtomCARE server of the KINS takes place via the SITS server located in the Emergency Operations Facility (EOF). However, the multi-path of transmission through SITS has caused problems such as an increase in data transmission interruptions and errors, as well as delays in identifying the cause and implementing system recovery measures. To address these issues, a new VPN is currently being constructed on the servers of nuclear facilities that generate and manage safety information to establish a direct transmission system of safety information from each nuclear facility to the AtomCARE server. The establishment of a direct transmission system that eliminates unnecessary transit steps is expected to result in stable information transmission and minimize the frequency of data transmission interruptions. As of the improvement progress, a security review was conducted in the second and third quarters of 2022 to evaluate the security of newly introduced VPNs to the nuclear facility server, and based on the results of the review, security measures were strengthened. In the fourth quarter of 2022, the development of a direct transmission system for safety information began, and it is scheduled to be completed by the fourth quarter of 2023. The project includes the construction of the transmission system, system inspection, and comprehensive data stability testing. Afterward, the existing SITS located in the EOF will be renamed as the Safety Information Display System (SIDS), and there are plans to remove any unused servers and VPNs.

Radionuclide analysis methods must be secured in the event of emergencies such as the discovery of unknown nuclear material or nuclear accidents in neighboring countries or Korea. Most institutions in Korea are in their early stages of radionuclide analysis method development and do not even have Radiation Controlled Areas where they can handle the samples safely. Some institutions such as the Korea Atomic Energy Research Institute have the ability to perform radionuclide analysis for nuclear facilities or verification of nuclear activities. In Korea, it is necessary to secure nuclide analysis technology to enable independent verification in times of emergency or need. This paper analyzes uranium as the target nuclide using alpha spectrometer and TIMS. Alpha spectrometer detects alpha particles emitted from uranium samples and measures the concentration of uranium isotopes. This method has a high selectivity that distinguishes it from other elements, and accurate measurements can be made even when uranium samples are mixed with other elements. In addition, there is minimal interference from other radioactive isotopes in the sample, and the sample preparation is simple, resulting in relatively short analysis times. In contrast, TIMS detects ionized uranium ions by heating the uranium sample. This method may have potential interference from other elements and may take relatively longer analysis times. However, TIMS has high sensitivity and accuracy and can detect various elements other than uranium, making it suitable for various analyses. Therefore, when analyzing uranium, it is recommended to select and use the appropriate device according to the purpose, as both alpha spectrometer and TIMS have their pros and cons. Furthermore, by using both devices in parallel, more accurate and reliable results can be obtained. This paper aims to compare the analysis methods of alpha spectrometer and thermal ionization mass spectrometry, which are widely used for nuclide analysis in unknown nuclear materials.

Employees of nuclear licensees have to take the education for radiological emergency preparedness, as prescribed by presidential Decree. The Korea Atomic Energy Research Institute (KAERI), as an educational institution designated by the Nuclear Safety and Security Commission (NSSC), has been conducting field-oriented workplace education. This aims to enhance understanding of radiological emergencies that may occur in nuclear facilities and to strengthen response capabilities to prevent and deal with accidents in the event of radiation emergencies or radioactive disasters. To accomplish these educational goals, a paradigm shift from the previous theory-oriented curriculum to a participatory curriculum with high field applicability is needed to strengthen the ability to respond to nuclear or radiological emergencies. In addition, a feedback system is required to manage the quality of education and improve the curriculum. In this regard, KAERI sought ways to revitalize the education to strengthen the emergency response competencies. Based on the concept of the Systematic Approach to Training (SAT) methodology, which is recommended by the International Atomic Energy Agency (IAEA) for the development and implementation of education and training for NPP personnel, an educational model and its feedback system were developed. Then, a field-oriented participatory curriculum operation and satisfaction survey were conducted to evaluate the educational effectiveness. Lastly, the survey results were discussed in a critique session to point out weaknesses and indicate areas for improvement, and then were used as data for educational quality assurance. This paper introduces the composition and effectiveness of KAERI’s SAT-based education model based on its recent three years of experience.

Radioactive materials depositied after nuclear accident or radiological emergency result in radiation exposure to individuals living in long-term contaminated territories. Therefore, the remedial actions should be taken on affected areas for the evacuated residents to return to their homes and normal lifestyle. Meanwhile, radiation exposure occurs through various pathways by work types during the site clean-up. Therefore, dose assessment is crucial to protect emergency workers and helpers from the potential radiological risk. This study estimated the exposure dose to individuals decontaminating the areas contaminated with 60Co, 63Ni, 90Sr, 134Cs, 137Cs, and then calculated the maximum workable soil concentration to comply with the reference level of 20 mSv/y for transition to existing exposure situations. For the realistic assessment, the detailed exposure scenarios depending on the types of work (excavation, collection, transportation, disposal, landfill), and the relevant exposure pathways were used. In addition, with the LHS (Latin Hypercube Sampling) - PRCC (Partial Rank Correlation Coefficient) method, sensitivity analysis was performed to identify the influence of the input parameters and their variation on the model outcomes. As a result, the most severe exposure-induced type was identified as the excavator operation with an annual individual dose of 4.75E-01 mSv at the unit soil concentration (1 Bq/g), from which the derived maximum workable soil concentration was 4.21E+01 Bq/g. Dose contribution by isotopes were found to be 60Co (55.63%), 134Cs (32.01%), and 137Cs (12.28%), and the impact of 63Ni and 90Sr were found to be negligible. Dose contribution by exposure pathways decreased in the following order: ground-shine, soil ingestion, dust inhalation, and skin contamination. Furthermore, the most high sensitive input parameters and their PRCC were found to be as the dilution factor (0.75) and as the exposure time (0.63). In conclusion, the results are expected to contribute to optimize radiation protection strategeis for recovery workers and to establish appropriate response procedures to be applicable in areas with high deposition density after a radiological or nuclear emergency.

As nuclear power plants are operated in Korea, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated. Due to the increase in the amount of radioactive waste generated, the demand for transportation of radioactive wastes in Korea is increasing. This can have radiological effect for public and worker, risk assessment for radioactive waste transportation should be preceded. Especially, if the radionuclides release in the ocean because of ship sinking accident, it can cause internal exposure by ingestion of aquatic foods. Thus, it is necessary to analyze process of internal exposure due to ingestion. The object of this study is to analyze internal exposure by ingestion of aquatic foods. In this study, we analyzed the process and the evaluation methodology of internal exposure caused by aquatic foods ingestion in MARINRAD, a risk assessment code for marine transport sinking accidents developed by the Sandia National Laboratory (SNL). To calculate the ingestion internal exposure dose, the ingestion concentrations of radionuclides caused by the food chain are calculated first. For this purpose, MARINRAD divide the food chain into three stages; prey, primary predator, and secondary predator. Marine species in each food chain are not specific but general to accommodate a wide variety of global consumer groups. The ingestion concentrations of radionuclides are expressed as an ingestion concentration factors. In the case of prey, the ingestion concentration factors apply the value derived from biological experiments. The predator's ingestion concentration factors are calculated by considering factors such as fraction of nuclide absorbed in gut, ingestion rate, etc. When calculating the ingestion internal exposure dose, the previously calculated ingestion concentration factor, consumption of aquatic food, and dose conversion factor for ingestion are considered. MARINRAD assume that humans consume all marine species presented in the food chain. Marine species consumption is assumed approximate and conservative values for generality. In the internal exposure evaluation by aquatic foods ingestion in this study, the ingestion concetration factor considering the food chain, the fraction of nuclide absorbed in predator’s gut, ingestion rate of predator, etc. were considered as influencing factors. In order to evaluate the risk of maritime transportation reflecting domestic characteristics, factors such as domestic food chains and ingestion rate should be considered. The result of this study can be used as basis for risk assessment for maritime transportation in Korea.

Radioactive waste can be classified according to the concentration level for radionuclides, and the disposal method is different through the level. Gamma analysis is inevitably performed to determine the concentration of radioactive waste, and when a large amount of radioactive waste is generated, such as decommissioning nuclear facilities, it takes a lot of time to analyze samples. The performance of a lot of analysis can cause human errors and workload. In general, gamma analysis is performed using by HPGe detector. Recently, for convenience of analysis, commercial automatic sample changers applicable to the HPGe detectors were developed. The automatic sample changers generate individual analysis reports for each sample. In this study, gamma analysis procedure was improved using the application of the automatic sample changer and the automated data parsing using by Python. The application of automatic sample changers and data parsing technique can solve the problems. The human errors were reduced to 0% compared to the previous method by improving the gamma analysis procedure, and working time were also dramatically reduced. This automation of analysis procedure will contribute to reducing the burden of analysis work and reducing human errors through various improvements.

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.