A high-efficiency underwater radiation monitoring system, HydroGamma, has been developed for detecting ¹³⁷Cs and ¹³¹I in the event of waterborne radiation contamination. The system consists of a 3-inch NaI (Tl) detector, solar panels for power supply, data acquisition and transmission modules, and batteries. HydroGamma also includes a ⁴⁰K calibration source for remote performance evaluation and energy calibration. In this study, some simulations and experiments were carried out to evaluate the minimum detectable activities (MDA) of HydroGamma. We installed the HydroGamma at Tapjeongho Lake in Nonsan-si and acquired background data since MDA is calculated based on the experimental background data. The results show that the minimum detectable activities for ¹³⁷Cs and ¹³¹I were 1.78Bq L^-1 and 1.81Bq L^-1, respectively even though the gamma rays emitted from ⁴⁰K(1,460 keV) affect the minimum detectable activities for them.

In this study, poly(ethylene-co-vinyl acetate)/magnesium hydroxide (EVA/MDH) composites were prepared by electron beam crosslinking. EVA as a matrix resin and MDH as a flame retardant were melt-blended and compression molded to prepare EVA/MDH composites. The prepared EVA/MDH composites were electron beam-irradiated at various absorbed doses of 50~200kGy. The effects of electron beam irradiation on the gel content, tensile strength, elongation-at-break, thermal properties, and flame retardancy of the composites were investigated. The gel content and tensile strength increased, while the elongation-at-break decreased with an increase in the absorbed dose due to the formation of crosslinked network structures. In addition, the thermal stability and flame retardancy improved as the absorbed dose increased. Therefore, the EVA/MDH composites prepared in this study can be used as an insulation material for flame-retardant and heat-resistant wires and cables.

This study was conducted to confirm the safety of the operator's radiation exposure in the micro PET-CT image acquisition experiment using the ¹⁸F-FDG. The usage of ¹⁸F-FDG and the exposure dose of handlers were measured at University B in Metropolitan City A, which uses ¹⁸F-FDG for micro PET-CT image acquisition. As a result of the measurement, the exposure dose is far below the effective dose limit of radiation workers, 50 mSv per year, and the equivalent dose limit of 500 mSv per year for hands, feet, and skin. has been measured Since these exposure doses can be further increased according to the number of times of use of ¹⁸F-FDG, it is judged that the exposure dose compared to the handling amount of ¹⁸F-FDG shown in this study can be used as reference data. In addition, as changed environments such as the use of materials other than unopened RI are occurring in education and research environments, such as the use of ¹⁸F-FDG at University B, radiation exposure with more interest in safety management by checking the factors of radiation exposure of the handler concerned We will always do our best to reduce it.

The classification of the radiation work performed in Korean nuclear power plants (NPPs) must be understood to provide workers with more comprehensive radiation protection. This study used annual reports on occupational exposure to investigate and analyze the similarities and differences in the radiation work performed in Korean NPPs with pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). The results showed that the radiation work performed in Korean NPPs could be classified into three categories. Category 1 contains work at the highest level. This work can be divided into individual tasks belonging to Category 2, which enables the evaluation of the radiation dose during the work. The work in Category 2 consists of tasks from Category 3, which contains basic detailed tasks that are not further subdivided. This study emphasized the need for the systematic management of the radiation work performed in both Korean PWRs and PHWRs, such as the tasks in Category 3, which are similar, with similar working conditions, for PWRs and PHWRs. It also suggested the need to establish a list of radiation work for decommissioning because Kori Unit 1 and Wolsong Unit 1 are currently in permanent shutdown and preparations are being made for their decommissioning.

This study was conducted to evaluate the applicability of the Comet assay, which is widely used for the identification of irradiated meats, to detect irradiated beef cut, ground beef, and Tteokgalbi during freezing storage. Gamma-irradiation significantly increased the DNA damage in frozen beef cut and ground beef samples. Among those, DNA nuclei of samples irradiated with absorbed doses of 1kGy or more showed typical comet-shaped damage, convincing that the samples were irradiated. Meanwhile, DNA nuclei in non-irradiated beef cut and ground beef samples were also damaged according to storage time. In particular, since the damage of DNA nuclei in the non-irradiated samples frozen for three months was similar to that of samples irradiated with a dose of 0.5 kGy, it was considered difficult to detect whether these samples were irradiated by Comet assay analysis. Likewise, gamma-irradiation of Tteokgalbi increased DNA damage. However, significant damage to DNA nuclei was observed even in the non-irradiated samples. Therefore, the application of the analysis method to determine whether the Tteokgalbi sample was irradiated was not appropriate. In conclusion, these results suggest that Comet assay could be limitedly applied only to fresh meat with a short storage period and minimal processing.

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

Workers in nuclear power plants are likely to be exposed to radiation from various geometrical sources. In order to evaluate the exposure level, the point-kernel method can be utilized. In order to perform a dose assessment based on this method, the radiation source should be divided into point sources, and the number of divisions should be set by the evaluator. However, for the general public, there may be difficulties in selecting the appropriate number of divisions and performing an evaluation. Therefore, the purpose of this study is to develop an algorithm for dose assessment for arbitrary shaped sources based on the point-kernel method. For this purpose, the point-kernel method was analyzed and the main factors for the dose assessment were selected. Subsequently, based on the analyzed methodology, a dose assessment algorithm for arbitrary shaped sources was developed. Lastly, the developed algorithm was verified using Microshield. The dose assessment procedure of the developed algorithm consisted of 1) boundary space setting step, 2) source grid division step, 3) the set of point sources generation step, and 4) dose assessment step. In the boundary space setting step, the boundaries of the space occupied by the sources are set. In the grid division step, the boundary space is divided into several grids. In the set of point sources generation step, the coordinates of the point sources are set by considering the proportion of sources occupying each grid. Finally, in the dose assessment step, the results of the dose assessments for each point source are summed up to derive the dose rate. In order to verify the developed algorithm, the exposure scenario was established based on the standard exposure scenario presented by the American National Standards Institute. The results of the evaluation with the developed algorithm and Microshield were compare. The results of the evaluation with the developed algorithm showed a range of 1.99×10^-1~9.74×10^-1 μSv hr^-1, depending on the distance and the error between the results of the developed algorithm and Microshield was about 0.48~6.93%. The error was attributed to the difference in the number of point sources and point source distribution between the developed algorithm and the Microshield. The results of this study can be utilized for external exposure radiation dose assessments based on the point-kernel method.

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.

The burnup profile of spent fuel should be determined accurately for the safety storage of spent fuel. In this study, a neutron detection system was developed as a part of basic research to analyze the burnup profile of spent fuel, and a performance was evaluated using a radiation source. The prototype of the neutron detection system was based on a ³He proportional chamber. The ³He proportional chamber is often used for neutron measurement and analysis because of its high neutron detection efficiency and simplicity for gamma ray rejection. For quantitative evaluation, tests were conducted using calibrated ²⁵²Cf and ¹³⁷Cs sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

Given that toxic substances are diffused by the various movements of the atmosphere, it is very important to evaluate the risks associated with this phenomenon. When analyzing the behavioral characteristics of these atmospheric diffusion models, the main input data are the wind speed and wind direction among the meteorological data. In particular, it is known that a certain wind direction occurs in summer and winter in Korea under the influence of westerlies and monsoons. In this study, synoptic meteorological observation data provided by the Korea Meteorological Administration were analyzed from January 1, 2012 to the end of August of 2022 to understand the regional wind characteristics of nuclear power plants and surrounding areas. The selected target areas consisted of 16 weather stations around the Hanbit, Kori, Wolsong, Hanul, and Saeul nuclear power plants that are currently in operation. The analysis was based on the temperature, wind direction, and wind speed data at those locations. Average, maximum, minimum, median, and mode values were analyzed using long-term annual temperature, wind speed, and wind direction data. Correlation coefficient values were also analyzed to determine the linear relationships among the temperature, wind direction, and wind speed. Among the 16 districts, Uljin had the highest wind speed. The median wind speed values for each region were lower than the average wind speed values. For regions where the average wind speed exceeds the median wind speed, Yeongju, Gochang, Gyeongju, Yeonggwang, and Gimhae were calculated as 0.69 m s^-1, 0.54m s^-1, 0.45m s^-1, 0.4m s^-1, and 0.36m s^-1, respectively. The average temperature in the 16 regions was 13.52 degrees Celsius; the median temperature was 14.31 degrees and the mode temperature was 20.69 degrees. The average regional temperature standard deviation was calculated and found to be 9.83 degrees. The maximum summer temperatures were 39.7, 39.5, and 39.3 in Yeongdeok, Pohang, and Yeongcheon, respectively. The wind directions and speeds in the 16 regions were plotted as a wind rose graph, and the characteristics of the wind direction and speed of each region were investigated. It was found that there is a dominant wind direction correlated with the topographical characteristics in each region. However, the linear relationship between the wind speed and direction by region varied from 0.53 to 0.07. Through this study, by evaluating meteorological observation data on a long-term synoptic scale of ten years, regional characteristics were found.

A multi-purpose canister (MPC) was developed for the purpose of transportation, storage and disposal of spent nuclear fuel (SNF) and has the advantage of minimizing repackaging between management stages of SNF. Considering the typical rock characteristics in Korea, a disposal canister is expected to contain 4 assemblies of Pressurized water reactor (PWR) SNF. The capacity of the MPC should be similarly designed with the disposal canister. However, the MPC with four SNF assemblies is expected to be less efficient in transporting and storing compared to a large-capacity canister. Therefore, a preliminary concept was derived for an auxiliary equipment that can transport and store multiple MPCs in a large overpack. A previously derived concept from US was thoroughly reviewed, and the preliminary concept was revised considering domestic situations including crane capacity and others. In addition, the safety of the normal transportation and storage of the MPC placed in transportation and storage overpack was evaluated with the auxiliary equipment.