To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

In the nuclear fuel cycle (NFC) facilities, the failure of Heating Ventilation and Air Conditioning (HVAC) system starts with minor component failures and can escalate to affecting the entire system, ultimately resulting in radiological consequences to workers. In the field of air-conditioning and refrigerating engineering, the fault detection and diagnosis (FDD) of HVAC systems have been studied since faults occurring in improper routine operations and poor preventive maintenance of HVAC systems result in excessive energy consumption. This paper aims to provide a systematic review of existing FDD methods for HVAC systems therefore explore its potential application in nuclear field. For this goal, typical faults and FDD methods are investigated. The commonly occurring faults of HVAC are identified through various literature including publications from International Energy Agency (IEA) and American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE). However, most literature does not explicitly addresses anomalies related to pressure, even though in nuclear facilities, abnormal pressure condition need to be carefully managed, particularly for maintaining radiological contamination differently within each zone. To build simulation model for FDD, the whole-building energy system modeling is needed because HVAC systems are major contributors to the whole building’s energy and thermal comfort, keeping the desired environment for occupants and other purposes. The whole-building energy modeling can be grouped into three categories: physics-based modeling (i.e., white-box models), hybrid modeling (i.e., grey-box models), and data-driven modeling (i.e., black-box models). To create a white-box FDD model, specialized tools such as EnergyPlus for modeling can be used. The EnergyPlus is open source program developed by US-DOE, and features heat balance calculation, enabling the dynamic simulation in transient state by heat balance calculation. The physics based modeling has the advantage of explaining clear cause-and-effect relationships between inputs and outputs based on heat and mass transfer equations, while creating accurate models requires time and effort. Creating a black-box FDD model requires a sufficient quantity and diverse types of operational data for machine learning. Since operation data for HVAC systems in existing nuclear cycle facilities are not fully available, so efforts to establish a monitoring system enabling the collection, storage, and management of sensor data indicating the status of HVAC systems and buildings should be prioritized. Once operational data are available, well-known machine learning methods such as linear regression, support vector machines, random forests, artificial neural networks, and recurrent neural networks (RNNs) can be used to classify and diagnose failures. The challenge with black-box models is the lack of access to failure data from operating facilities. To address this, one can consider developing black-box models using reference failure data provided by IEA or ASHRAE. Given the unavailability of operation data from the operating NFC facilities, there is a need for a short to medium-term plan for the development of a physics-based FDD model. Additionally, the development of a monitoring system to gather useful operation data is essential, which could serve both as a means to validate the physics-based model and as a potential foundation for building data-driven model in the long term.

With an ultimate view to identifying abnormal releases of radioactive materials, a set of liquid and gaseous effluent data including unplanned or uncontrolled releases annually reported form the U.S. and Korean nuclear power plants were systematically analyzed. With the use of 21 years’ worth of annual discharge data for 7 radionuclide groups and 24 individual radionuclides, taken from a combined total of 1,610 reactor-years (RYs) covering 62 units of US Pressurized Water Reactors (PWRs) and 22 units of Korean PWRs, three novel formulas for estimating events were employed to calculate characteristic values. Applying these characteristic values derived from the event estimation formulas to events that transpired during 699 RYs in operational US PWRs revealed an enhanced predictive accuracy for abnormal events when considering individual radionuclides, as opposed to grouping them by radionuclide groups. This effect was particularly pronounced for specific events such as leaks caused by problems in Gas Decay Tanks, leaks in Steam Generator Power Operated Relief Valves, fuel defects, and leaks during spent nuclear fuel processing. In the case of Korean PWRs, fuel defects were identified as the primary events related to radioactive effluent releases. The methodologies and characteristic values derived from this study were applied to these events. The event estimation rate was lower in Korean compared to US PWRs, which can be attributed to the lower frequency of event occurrences in Korean PWRs (30 RYs) compared to the US. The approach proposed in this study may contribute to develop a methodology to identify implicit abnormal release data and correlate them with specific operational occurrences or events, which could improve the conventional practice of simply recording and reporting radioactive discharge data.

If radioactive plumes are released outside due to loss of containment building integrity during a nuclear power plant accident, these materials might travel with the wind, affecting both the surrounding environment and neighboring countries. In China, most nuclear power plants are located on the eastern coast. Consequently, a radioactive plume generated during an accident could negatively impact even the western part of the Korean Peninsula due to westerly winds. To detect such problems early, respond quickly, and protect residents, a system that can monitor aerial radiation under normal conditions is needed. Additionally, a detection system that can operate in real-time in an emergencies conditions is required. The current method for aerial radiation measurement takes environmental radiation data from a monitoring post 1.5 m above the ground and converts it to altitude. To measure actual aerial radiation, an expansive area is surveyed by aircraft. However, this approach is both time-consuming and expensive. Thus, to monitor radioactive plumes influenced by environmental factors like wind, we need a radiation detector that can gauge both radioactivity and directionality. In this study, we developed a radiation detector capable of assessing both the radioactivity and directionality of a radioactive plume and conducted its performance evaluation. We miniaturized the radiation detector using a CZT (Cadmium Zinc Telluride) sensor, enabling its mounting on unmanned aerial vehicles like drones. It is configured with multi-channels to measure directionality of a radioactive plumes. For performance evaluation, we positioned two-channel CZT sensors at 90 degrees and measured the energy spectrum for angle and distance using a disk-type radioactive isotope. Using this method, we compared and analyzed the directionality performance of the multi-channel radiation detector. We also confirmed its capability to discern specific radioactivity information and nuclide types in actual radioactive plumes. Our future research direction involves mounting the multi-channel radiation detector on a drone. We aim to gather actual aerial radiation data from sensors positioned in various directions.

To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

Detectors utilized for nuclear material safeguards have been using scintillation detectors which are inexpensive and highly portable, and electrically cooled germanium detectors which are expensive but have excellent energy resolution. However, recently IAEA, the only international inspectorate of nuclear material safeguards for the globe, have replaced the existing scintillation detector and electrically cooled germanium detector with a CdZnTe detector owing to the improved performance of room-temperature semiconductors significantly. In this paper, we will examine the spectrum features of the CdZnTe detector such as spectrum shape, energy resolution, and efficiency in the energy region of interest, which are the important characteristics for measuring Uranium enrichment. For this purpose, it would be conducted to compare its spectrum features using CdZnTe, NaI, HPGe detectors. The main energies of interest include 185.7 keV and 1,001 keV, which are the decay energies of uranium 235 and uranium 238. The results of this study will provide a better understanding of the spectral features of various detectors used in uranium enrichment analysis, and are expected to be used as basic data for future related software development.

Radiation workers, especially those dealing with Uranium isotopes, can potentially intake Uranium -containing materials through their respiratory and digestive systems. According to the “Regulations on the Measurement and Calculation of Internal Exposure” from Nuclear Safety and Security Commission (NSSC), those who intend to work in or enter the nuclear facilities with a risk of exceeding 2 mSv exposure per year should be examined the internal exposure. However, when it comes to in-vitro bioassay, Uranium intake through drinking water can affect the quantitative analysis. The International Commission on Radiological Protection (ICRP) reported in ICRP Publication 23 (Report on the Task Group on Reference Man) that the reference man excretes Uranium in the urine (0.05-0.5 μg/day) and feces (1.4-1.8 μg/day). Korea Atomic Energy Research Institute (KAERI) set the 90.5 ng/day as the 238U background of workers handing Uranium based on the daily Uranium intake of Koreans. In this research, we examined the possible effects of Uranium in drinking water on internal exposure by analyzing the concentration of Uranium in bottled waters from various water sources sold in the domestic market and a water from the water purifier. The 238U concentration results of analyzing 11 bottled waters and 1 purified water, were ranged from 0 to 10.2 μg/L. All the results were satisfied the standard of 30 μg/L according to “Regulations for Drinking Water Quality Standards and Inspection” enacted by the Ministry of Environment. However, various concentrations were shown depending on the water sources. Assuming that these concentrations of water are consumed by drinking 1 L per day, the internal dose assessment result is 0 to 0.94 mSv. On the other hand, if it is assumed to be inhaled, it can be an overestimated because the dose coefficient of inhalation, Type M is higher than that of ingestion, f1=0.02 which are the values recommended by ICRP Publication 78 (Individual Monitoring for Internal Exposure of Workers) when the Uranium compound is unspecified. In case of two workers at KAERI, the daily excretion of urine was 151 and 120 ng/day respectively in the first quarter monitoring. However after changing the kind of drinking water in the second quarter monitoring, it dropped to 17.4 and 15.4 ng/day respectively. Through this study, it is confirmed that the Uranium background in urine can be analyzed differently depending on the kind of drinking water consumed by each worker. Depending on the Uranium concentration of drinking water, the internal exposure dose assessment can be overestimated or underestimated. Therefore, the Uranium concentration and intake amount according to the kind of drinking water should be considered for in-vitro bioassays of Uranium handlers. Furthermore, if necessary, the Uranium isotope ratio analysis in urine and the handling information should be comprehensively considered. In addition, in order to exclude the effect of intake through the digestive system, replacing the kind of drinking water can be considered. The additional analysis such as in-vivo bioassay and 24 hours urine analysis rather than spot samples can be also recommended.

The demand for transportation is increasing due to the continuous generation of radioactive wastes. Especially, considering the geographical characteristics of Korea and the location characteristics of nuclear facilities, the demand for maritime transportation is expected to increase. If a sinking accident happens during maritime transportation, radioactive materials can be released into the ocean from radioactive waste transportation containers. Radioactive materials can spread through the ocean currents and have radiological effects on humans. The effect on humans is proportional to the concentration of radioactive materials in the ocean compartment. In order to calculate the concentration of radioactive materials that constantly flow along the ocean current, it is necessary to divide the wide ocean into appropriate compartments and express the transfer processes of radioactive materials between the compartments. Accordingly, this study analyzed various ocean transfer evaluation methodologies of overseas maritime transportation risk codes. MARINRAD, POSEIDON, and LAMER codes were selected to analyze the maritime transfer evaluation methodology. MARINRAD divided the ocean into two types of compartments that water and sediment compartments. And it was assumed that radionuclides are transfered from water to water or from water to sediment. Advection, diffusion, and sedimentation were established as transfer process for radionuclides between compartments. MARINRAD use transfer parameters to evaluate transer processes by advection, diffusion, and sedimentation. Transfer parameters were affected by flow rate, sedimentation rate, sediment porosity, and etc. POSEIDON also divided the ocean into two types that water and sediment compartment, each compartments was detaily divided into three vertical sub-compartment. Advection, diffusion, resuspension, sedimentation, and bioturbation were established as transport processes for radionuclides between compartments. POSEIDON also used transfer parameters for evaluating advection, diffusion, resuspension, sedimentation, and bioturbation. Transfer parameters were affected by suspended sediment rates, sedimentation rates, vertical diffusion coefficients, bioturbation factors, porosity, and etc. LAMER only considered the water compartment. It divided the water compartment into vertical detailed compartments. Diffusion, advection and sedimentation were established as the nuclide transfer processes between the compartments. To evaluated the transfer processes of nuclides for diffusion and advection, LAMER calculated the probability with generating random position vectors for radionuclides’ locations rather than deterministic methods such as MARINRAD’s transfer parameters or POSEIDON’s transfer rates to evaluate transfer processes. The results of this study can be used as a basis for developing radioactive materials’ ocean transfer evaluation model.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

The radioactive cesium, released from the normal operation or the accidental operation of nuclear facilities, should be regularly monitored for environmental regulatory compliance. The 135Cs/137Cs isotopic ratios, potentially useful for long-term tracking Cs transport in seawater, can be used as a tool of understanding how radionuclides are transported from different nuclear production source terms and distributed in the ocean. The ultra-high sensitive mass spectrometers (TIMS, SF-ICP-MS and TQ-ICP-MS) have been used to measure the 135Cs/137Cs isotopic ratios. However, the radiochemical separation of Cs from the seawater matrix is essential for the analysis of Cs using the mass spectrometers. An automated radiochemical procedure for the separation of Cs in seawater was developed for the analysis of 135Cs/137Cs isotopic ratios using a sequential column chromatography with AMPPAN and AG50Wx8 cation exchange resins. National Instrument’s LabVIEW is a graphical programming language and a powerful tool for the instrument control. A virtual instrument system for the automated separation of cesium isotopes was developed by the state machine of the fundamental design patterns in LabVIEW. In this study, the conceptual designs of an automated separation system of cesium isotopes, its virtual instrument system based on the LabVIEW state machine architectures and an automated radiochemical procedure were described for the purification of cesium isotopes at trace levels found in seawater discharged from the various nuclear facilities.

South Korea’s first commercial nuclear reactor, Kori Unit 1, was permanently shut down in 2017, and preparations are currently underway for its decommissioning. After the permanent shutdown, the spent nuclear fuel from the reactor core is removed and stored in a spent fuel storage facility. Subsequently, steps are taken for its permanent disposal, and if a permanent disposal site is not determined, it is stored in an interim storage facility (or temporary storage facility). Therefore, the activation criteria for radiation emergency plans vary depending on the movement of spent nuclear fuel and the storage location. In this study, it reviewed emergency plans in the U.S. NRC Regulatory Guide (Draft) titled ‘Emergency Planning for Decommissioning Nuclear Power Reactors’ to determine the requirements for radiation emergency plans needed for decommissioned nuclear power plants. Additionally, by examining emergency plans applied to decommissioning nuclear power plants in the United States, this study identified emergency plan requirement that could be applicable to future decommissioned nuclear power plants in South Korea. This study will contribute to the establishment of appropriate radiation emergency plans for decommissioning nuclear power plants in Korea for providing accurate information on overseas cases and relevant guidelines.

When occurring at a nuclear power plant (NPP) by accidents, accurate prediction and identification of the process of radioactive material dispersing into atmosphere is important to protect public and environment. Atmosphere dispersion of radioactive materials is significantly influenced by wind direction and wind speed. The government and nuclear operator continuously monitor wind data at nuclear sites through meteorological tower to prepare for such accidents involving the release of radioactive materials. The purpose of this study is to construct wind rose diagrams at 5 NPP sites (Kori, Saewool, Wolsong, Hanbit, Hanul). Wind roses serve as invaluable tool for identifying wind patterns in each region and visualizing wind directions. This can be utilized to predict the dispersion pathway and extent range of radioactive materials carried by the wind. This program will take on the role of establishing appropriate evacuation routes or shelter locations for residents when reliable wind data is not immediately available during an NPP accident. The wind data used in the study was collected from a meteorological tower located at the NPP site, and measurements were taken at 1-hour intervals for each operation over a period of ten years. The collected data underwent preprocessing, followed by the development of Python code to render the wind rose diagrams in an interpretable format. The future direction of this study will be focused on enhancing this program by integrating geographical mapping capabilities. With these advancements, it will become feasible to superimpose shelter positions on a map in accordance with prevailing wind directions. These improvements will contribute to the development of additional protective measures for residents and the proposal of alternative shelter options in response to potential radioactive material releases.

According to IAEA GSR Part.6, Decommissioning is carried out on the basis of planning and evaluation to ensure safety, protection of workers, public, and environment. Then, the decommissioning project of nuclear facility includes a radiation protection plan that reflects the regulatory requirements and international recommendations of each country and the internal regulations of the licensee. The scope of the radiation protection plan covers all radiation activities related to the dismantling and disposal of contaminated facilities subject to decommissioning. Radiation protection applications in the United States, a country with previous experience in decommissioning nuclear facilities, include 10 CFR 20 for NRC management facilities and 10 CFR 835 for facilities under DOE. In this study, we analyzed two cases of decommissioning plans to which NRC regulations are applied. In 1992, Yankee Atomic Electric Company (YAEC), the licensee of Yankee Nuclear Power Station (YNPS), notified NRC of the permanent shutdown of YNPS and submitted decommissioning plan accordingly. This decommissioning plan consists of a total of 9 chapters, and section 3.2 describes the radiation protection of decommissioning workers. The contents of the radiation protection program consist of 16 subsections. Another case is the decommissioning work plan of U.S. Navy Surface Ship Support Barge (SSSB), which used in Virginia to support the refueling of the U.S. Navy’s reactor vessel. This document was developed based on the NUREG-1757 and was revised in 2021 after receiving NRC comment. SSSB’s project radiation protection plan is described in appendix 1, and the contents consist of a total of 28 sections except for reference. In Korea, decommissioning plan is developed in accordance with “Standard Format and Content of the Decommissioning Plan for Nuclear Facilities”. According to this regulation, the radiation protection plan for licensing documents submitted at the time of application for approval of decommissioning execution shall describe the organization and functions for implementing of plan, methods, cycles and procedures for performing radiation protection and radiological monitoring. Also, the safety review guidelines of regulatory body also require radiation protection plans and procedures to ensure ALARA activities during decommissioning. In the case of the final decommissioning plan of Kori-1, which is currently submitted to regulatory body for licensing review, the decommissioning radiation protection plan is divided into 8 sections. Although the classification criteria for the radiation protection plan categories described above facilities are different, it could be seen that the following 7 contents are included in common: (a) ALARA application and organization for implementation, (b) Management of radiation control area, (c) Process of radiation work, (d) Radiation and contamination control, (e) Personnel radiation exposure monitoring, (f) Radioactive material management, (g) Radiation protection training.

A new annual dose evaluation system called E-DOSE has been developed. The system is based on the methodology of the previous version, K-DOSE60, which uses the dose evaluation methods of the International Commission on Radiological Protection (ICRP-60). However, E-DOSE is coded in ABAP to be compatible with the KHNP’s enterprise resource planning (ERP) system, SAP. This allows E-DOSE to use the real-time data from SAP, which minimizes the need for user intervention. The socio-environmental data, which was previously managed by the staff of each plant sites, can now managed in the system in a centralized manner. This is a significant improvement over the previous system, as it reduces the risk of errors and makes it easier to track and manage data. The system also automatically generates the reports required by regulations. EDOSE is expected to minimize the occurrence of human errors in preparing and managing the input data. This is because the system uses the data from SAP, which is less prone to errors than manually entered data. Additionally, the automatic generation of reports reduces the risk of errors in report preparation. E-DOSE is also expected to improve work efficiency. This is because the system automates many of the tasks involved in annual dose evaluation, such as data entry, calculation, and report generation. Overall, E-DOSE is a significant improvement over the previous annual dose evaluation system. It is more efficient, accurate, and user-friendly.

The nuclear licensee must ensure that the nuclear or radiological emergency preparedness and response organization is explicitly defined and staffed with adequate numbers of competent and assessed personnel for their roles. This paper describes the responsibilities of medical and support personnel for the medical action of casualties in the event of a radiological emergency at the KAERI. Currently, there is one medical personnel (nurse) in KAERI, and a total of eight medical support personnel are designated for medical response in the event of a radiological emergency. These medical support personnel are designated as one or two of the on-site response personnel at each nuclear facility, operating as a dedicated team of A, B (4 people each). In the event of a radiological emergency, not all medical support personnel are mobilized, but members of the dedicated medical team, which includes the medical support personnel of the nuclear facility where the accident has occurred, are summoned. Medical and support personnel will first gather in the onsite operational support center (OSC)/technical support center (TSC) to prepare and stand by for the medical response to injured when a radiological emergency is declared. They should take radiation protective measures, such as wearing radiation protective clothing and dosimeters, before entering the onsite of a radiological emergency, because injuries sustained during a radiological emergency may be associated with radioactive contamination. In the event of an injury, direct medical treatment such as checking the patient’s vitals, first aid, and decontamination will be carried out by medical personnel, while support personnel are mainly responsible for contacting the transfer hospital, reporting the patient’s condition, accompanying the ambulance, filling out the emergency medical treatment record, and supporting medical personnel. In order to respond appropriately to the occurrence of injuries, we regularly conduct emergency medical supplies education and medical training for medical support personnel to strengthen their capabilities.

In the event of a radiological emergency at a nuclear facility, the exchange of information on the accident situation is very important in the response process. For this reason, international organizations such as the IAEA and the EU operate systems to exchange information in the event of a radiological emergency. In south korea, the emergency response information exchange system (ERIX) developed by KINS is operated for use by the national radiological emergency response organization. The ERIX enables the exchange of emergency response information between organizations such as the government, nuclear operators, local authorities, KINS and KIRAMS. The KAERI has developed the KAERI emergency response information exchange system (KAERIX) for the exchange of accident information and emergency response information between the emergency response organizations of the KAERI in the event of a radiological emergency. This system is web-based using HTML, runs on internal network and is only available to KAERI staff. Recently, as the need for optimizing and upgrading KAERIX has arisen, improvements have been derived. The main improvement is optimizing KAERIX for Microsoft Edge to minimize errors. At present, it is optimized for Internet Explorer, but optimizing it for Microsoft Edge mode has become essential due to Microsoft discontinuing support for Internet Explorer. Another major improvement involves adding functions in ERIX to KAERIX, such as displaying the deletion/ correction status of input information and providing notifications for important information registration. Ultimately, KAERIX will be upgraded and optimized in 2024, reflecting improvements.

Domestic nuclear power plants conduct radiological environmental impact assessments every year in accordance with the Nuclear Safety and Security Commission (NSSC) notice. Among them, gaseous effluents are evaluated for their effects due to inhalation, external exposure in the air, exposure from ground surface deposits, food intake. In order to evaluate the impact of this exposure pathway, an evaluation point for each pathway must be selected. In the case of evaluation points, each country has different evaluation points. In the case of Korea, the evaluation point is calculated on the assumption that one lives 365 days a year at the EAB and consumes food from the nearest production area. In the case of the United States, external exposure and inhalation are evaluated at the site boundary or the nearest residential area, and food intake is evaluated by assuming that food produced in the nearest residential area or the nearest production area is consumed. Currently, the dose evaluation is optimized and selected so that EAB evaluation point for each site includes 16 direction evaluation points for each unit. In the E-DOSE60 program currently under development, the evaluation point was selected by calculating 16 direction x number of units without optimization. The food intake evaluation point was selected as the point that satisfies the minimum farmland area of the U.S. NRC NUREG-1301 and is the shortest distance from the site. The location of the production point from multiple units in included all 16 directions for each unit and quantity of evaluation points was optimized to satisfy the shortest distance. It can contribute to improving the reliability of the E-DOSE60 program currently under development by selecting new evaluation points for evaluating inhalation and external exposure evaluation and selecting optimized dose evaluation points for each site for evaluation by ingestion.

Korea Atomic Energy Research Institute (“KAERI”) has been developing various studies related to the nuclear fuel cycle. Among them, KAERI was focusing on the pyroprocess, which recycles some useful elements white reducing the volume and toxicity of spent nuclear fuel (SNF). Pyroprocess involves the handling of SNF, which cannot be handled directly by the facility worker. Therefore, SNF is handled and processed through remote handling device within a shielded facility such as a hot cell. Nuclear Facilities with such hot cells are called nuclear fuel cycle facilities, and unlike other facilities, heating, ventilating, and air conditioning (HVAC) system are particularly important in nuclear fuel cycle facilities to maintain the atmosphere in the hot cell and remove radioactive materials. In addition, due to the nature of the pyroprocess, which uses molten salt, corrosion is a problem in air atmosphere, so the process can only be carried out in an inert gas atmosphere. KAERI has a nuclear fuel cycle facility called the Irradiation Material Examination Facility (IMEF), and has built and operated the ACPF inside the IMEF, which operates an inert atmosphere hot cell for the demonstration of the pyroprocess. For efficient process development of the pyroprocess, it is necessary to put the developed equipment into the hot cell, which is a radiationcontrolled area, after sufficient verification in a mock-up facility. For this purpose, the ACPF mock-up facility, which simulates the system, space, and remote handling equipment of the ACPF, is operated separately in the general laboratory area. The inert gas conditioning system of the ACPF consists of very complex piping, blowers, and valves, requires special attention to maintenance. In addition, if there is a small leak in the piping within these valves or piping, radioactive materials can be directly exposed to facility workers, so continuous monitoring and maintenance are required to prevent accident. In this study, the applicability of acoustic emission technology and ultrasonic technology for leak detection in the inert gas conditioning system of ACPF mock-up facility was investigated. For this purpose, new bypass pipes and valves were installed in the existing system to simulate the leakage of pipes and valves. Acoustic emission sensors are attached directly to pipes or valves to detect signals, while ultrasonic sensors are installed at a distance to detect signals. The optimal parameters of each technology to effectively suppress background noise were derived and, and the feasibility of identifying normal and abnormal scenarios in the system was analyzed.

Korea Atomic Energy Research Institute (“KAERI”) has been developing pyroprocess technology for the sustainable use of nuclear energy and radioactive waste reduction, and is conducting design studies for a Pyroprocess Commercializing Research Facility (PCRF). High-level radioactive materials such as spent nuclear fuel, which are handled in the hot cell of the PCRF, physically change materials directly or cause chemical changes through ionization or excitation depending on the energy and types of radiation. Therefore, all facilities, including process equipment and remote handling equipment, installed into the hot cell must be evaluated for radiation hardness to be maintained in the radiological environmfent so that processes can proceed throughout the design life of the facility. In addition, as the maintenance paradigm has recently shifted from corrective maintenance to predictive maintenance, it is necessary to know in advance the condition of the equipment or facility in the radiological environment. In this study, an analysis of the radiation environment of the hot cell in the PCRF was conducted through source term, and the radiological dose impact was evaluated through the results of irradiation experiments of major components by reference data. Then, the actual dose contribution was identified through dose assessment using the MCNP code based on the pyroprocess equipment, and the radiation hardness requirements for the facility and equipment in the hot cell were derived by the above results.

When decommissioning a nuclear power plant, it is expected that clearance or radioactive waste (e.g., soil, concrete, metal, etc.) below the low-level will be generated in a short period on a large scale. Among the various types of waste, most of the contaminated soil is known to be classified as clearance or the (very) low-level radioactive waste. Accordingly, an accurate measurement and classification of contaminated soil in real-time during the decommissioning process can efficiently reduce the amount of soil waste and the possibility of contamination diffusion. However, in order to apply a system that measures and classifies contaminated soil in real-time according to the level of contamination to the decommissioning site, a demonstration is required to evaluate whether the system is applicable to the site. In this study, to establish requirements for determining the applicability of the system to the decommissioning site, preceding cases from countries with abundant decommissioning experience were investigated. For example, MACTEC of the U.S. demonstrated the developed system at the Saxton nuclear power plant in the U.S. and confirmed that the amount of soil that can be analyzed per hour in the system is affected by radionuclides, minimum detectable activity (MDA), and applicable volume. In the future, therefore, we will utilize the result of this study to develop the requirements of demonstrating the system for measurement and classification of contaminated soil in real-time.