In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

Concrete is the primary building material for nuclear facilities, making it one of the most common forms of radioactive waste generated when decommissioning a nuclear facility. Of the total waste generated at the Connecticut Yankee and Maine Yankee nuclear power plants in the United States, concrete waste accounts for 83.5% of the total for Connecticut Yankee and 52% for Maine Yankee. In order to dispose of the low- to medium-level radioactive concrete waste generated during the decommissioning of nuclear power plants, it is necessary to analyze the radioactivity concentration of gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144. Gamma-ray spectroscopy is commonly used method to measure the radioactivity concentration of gamma nuclides in the radioactive waste; however, due to the nature of gamma detectors, gamma rays from sequentially decaying nuclides such as Co-60 or Y-88 are subject to True Coincidence Summing (TCS). TCS reduces the Full Energy Peak Efficiency (FEPE) of specific gamma ray and it can cause underestimation of radioactivity concentration. Therefor the TCS effect must be compensated for in order to accurately assess the radioactivity of the sample. In addition, samples with high density and large volume will experience a certain level of self-shielding effect of gamma rays, so this must also be compensated for. The Radioactive Waste Chemical Analysis Center at the Korea Atomic Energy Research Institute performs nuclide analysis for the final disposal of low- and intermediate-level concrete waste. Since a large number of samples must be analyzed within the facility, the analytical method must simultaneously satisfy accuracy and speed. In this study, we report on the results of evaluating the accuracy of the radioactivity concentration correction by applying an efficiency transfer method that appears to satisfy these requirements to concrete standard reference material.

In order to establish disposal plans for sludge, which is one of the untreated waste materials from domestic nuclear power plants, it is necessary to determine the radioactivity concentration of radioactive isotopes. In this study, we aim to evaluate the gross alpha radioactivity of sludge containing radioactive contaminants after pre-treatment, in order to assess the level of sludge waste and obtain analytical data for discussing disposal methods. Samples of sludge generated from nuclear power plants were pre-treated, solutionized, and prepared as analysis samples for evaluating the gross alpha radioactivity.

Radiation from spent nuclear fuel (SNF) is one of key factors affecting the dissolution process of SNF and the source term from repository. The dissolution rate of uranium dioxide (UO2) matrix of SNF is expected to control the release of radionuclides from SNF in contact with water under geological disposal conditions. Based on the oxidative dissolution mechanism, the solubility of UO2 can increase significantly if the reducing environment near the fuel surface is altered by water radiolysis caused by radiation from SNF. Therefore, the analysis of water radiolysis products such as radicals (·OH, ·OH2, eaq, ·H) and molecules (H3O+, H2, H2O2) is perquisite for studies on the rate of such dissolution process to determine oxidation/dissolution mechanism and related rate constants. In this study we examined the two-known spectroscopic methods developed for H2O2 determination; one is the luminol-based chemiluminescence (luminol-CL) method and the other is the spectrophotometry using ferrous oxidation-xylenol orange complexation (FOX). Their applicability for quantitative analysis of H2O2 in potential aqueous samples from SNF dissolution studies was evaluated in terms of the analytical dynamic range (ADR), the limit of detection (LOD) and the interfering effects of various metal ions possibly present in real samples. The luminol-CL method exploits the chemiluminescence reaction caused by luminol; when in the presence of a metallic catalyst (e.g., Cu2+, Co2+), luminol emits a blue light (425 nm) at pH 10- 11 in response to oxidizing agents such as hydrogen peroxide. Although a flow-through reaction system is routinely employed to enhance the analytical sensitivity we achieved the ADR up to ~200 μM and LOD < 1 μM by a batch-wise CL detection using conventional cuvette cells and an intensified charge-coupled device (ICCD). Interestingly, it turned out that the interfering effects of other metal ions (e.g., UO2 2+, U4+, Fe2+ and Fe3+) is minimal, which should be advantageous for the luminol-CL method to be employed for samples potentially containing other metal ions. On the other hand, the FOX method spectrophotometrically analyzes H2O2 based on the difference in color (or absorption spectra) of Fe-xylenol orange (XO) complexes. Initially, the Fe2+-XO complex was provided in working solutions at pH 3, which was subsequently mixed with samples having H2O2 and allowed for quantitative oxidation of Fe2+ to Fe3+. Typically, by monitoring the absorbance of Fe3+-XO complex at 560-580 nm (λmax) the ADR up to ~100 μM and LOD ~1.6 μM were achieved. However, it is found that interfering effects from M3+ and M4+ ions are significant; these interfering metal ions can form XO complexes so as to directly contribute the measured absorbance. In contrast, the influence from M2+ ions was found to be negligible. To summarize we conclude that both methods can be applied for H2O2 determination for aqueous samples taken from SNF dissolution tests. However, prior to applying the FOX method the metal ion composition in those samples should be thoroughly identified not to overestimate the H2O2 concentration of samples. More details of underlying chemical reactions in both methods will be discussed in the presentation.

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.

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 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.

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.

Various types of radioactive liquid and solid wastes are generated during the operation and decommissioning of nuclear power plants. To remove radionuclides Co-60, Cs-137 etc. from a liquid waste, the ion-exchange process based on organic resins has been commonly used for the operation of nuclear facilities. Due to the considerations for the final disposal of process endproduct, other treatment methods such as adsorption, precipitation using some inorganic materials have been suggested to prepare for large amounts of waste during decommissioning. This study evaluated sintering characteristics for radioactive precipitates generated during the liquid waste treatment process. The volume reduction efficiency and compressive strength of sintered pellets were the major parameters for the evaluation. Major components of a simulated precipitate were some coagulated (oxy) hydroxides containing light elements, such as Si, Al, Mg, Ca, and zeolite particles. Green pellets compressed to around 100 MPa were heated at a range of 750~850°C to synthesize sintered pellets. It was observed that the volume reduction percentages were higher than 50% in the appropriate sintering conditions. The volume reduction was caused by the reduction of void space between particles, which is an evidence of partial glassification and ceramization of the precipitates. This result can also be attributed to conversion reactions of zeolite particles into other minerals. The compressive strength ranged from 6 to 19 MPa. These results also showed a significant correlation with the volume reduction of sintered body. Although our lab-scale experiments showed many benefits of sintering for the precipitates, optimized conditions are needed for large-scale practical applications. Evaluation of sintering characteristics as a function of pellet size and further testing will be conducted in the future.

Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

In South Korea, the replacement of steam generators began with Kori Unit 1 in 1995, and to date, 20 steam generators have been replaced and are currently stored in intermediate storage facilities. In the future, additional decommissioned steam generators may arise due to measures like the extension of the lifespan of nuclear power plants. In Korea, technological development for dismantling steam generators is underway, and there is no track record of actual dismantling. Although the replaced decommissioned steam generators are stored in intermediate facilities, for site recycling purposes, steam generators, which have relatively lower radiation doses compared to reactor heads and other primary equipment, should be prioritized for dismantling. While there are various specifications for steam generators, those dismantled and stored domestically are of the Recirculation Type. They can be classified into three types: the Westinghouse type WH-51 used in Kori Unit 1, the Fra-51B used in Han-ul Units 1 and 2, and the OPR-1000 used in Han-ul Units 3 and 4. The quantity of U-Tubes varies depending on the specification, but the radiation is concentrated in the primary side components, the U-Tube and Chamber. Since the parts related to the secondary side are not contaminated, they can be disposed of independently after classification. To dismantle a steam generator, it is of utmost importance to first create a scenario regarding where and how the dismantling will take place. Through the analysis of the advantages and disadvantages of each scenario, the optimal timing, location, and cutting method for dismantling should be researched. Furthermore, based on those findings, the best scenario should be derived through an analysis of worker radiation exposure and dismantling costs. To achieve this, a 3D simulation software developed by Cyclelife Digital Solutions under the French EDF was utilized to conduct simulations based on different dismantling schedules and methods. As a result, the optimal scenario for dismantling the steam generator was derived.

In the ocean, there exist infinite resources, including certain metallic elements that can serve as potential energy sources. One of the methods for extracting these dissolved resources from seawater involves adsorption. This study discusses the results of experiments conducted in real seawater using a developed fiber-type adsorbent capable of extracting dissolved oceanic resources. The fiber-type adsorbent was deployed in seawater to adsorb the elemental resources. It was then retrieved after 2, 3, and 4 weeks for evaluation of its adsorption performance. The evaluation was carried out by dissolving the adsorbent in a strong acidic solution and calculating the adsorption amount per gram of adsorbent using ICP-MS. The results indicated that the adsorption performance was slightly lower than previously reported values. Nevertheless, it confirmed the feasibility of adsorbing and recovering dissolved resources from actual seawater

Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes generated from nuclear power plant decommissioning. It is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. The contamination evaluation module of the system has two sub modules. One is for quick measurement with NaI (Tl) detector and the other is for accurate measurement with HPGe detector. The container used at the system for wastes handling has capacity of 100 kg and made of stainless steel. According to the measurement result of Co-60 and Cs-137, the waste is classified as waste for disposal or waste for clearance. Performance of the system was demonstrated using RM (Reference Material) radiation source. This year, necessity of system improvement was suggested due to revised operation requirements. So, the system should show throughput of more than 1 ton/hr and Minimum Detectable Activity (MDA) of less than 0.01 Bq/g (1/10 of criteria for regulatory clearance) for Co-60 and Cs-137. And soil waste become main target of the system. For this, the container used for soil waste handling should have capacity of 200 kg. As a result, material for the container need to be changed from stainless steel to plastic or FRP (Fiber Reinforced Plastics). And large area detector should be introduced to the system to enhance processing speed of the system. Additionally, container storage rack and conveyor system should be modified to handle 200 kg capacity container. Finally, moving path of the container will be redesigned for enhanced throughput of the system. In this paper, concept development of the system was suggested and based on that, system development will be followed.

When dismantling a power plant, a large amount of radioactive tanks are generated, and it is estimated that a significant amount of sludge will accumulate inside the tanks during long-term operation. In the process of dismantling a radioactive tanks, it is important to know the composition of the sludge because the sludge present inside must first be removed and then disposed of. In the case of certain tanks, it can be predicted that corrosion products generated due to system corrosion are the main cause of sludge formation. However, in the case of some tanks, it is not easy to predict the sludge composition because various dispersed particles in addition to corrosion products may be mixed with the wastewater. Even if it is collected and analyzed, the sludge composition can change significantly depending on the operation history, so the analysis results cannot be considered representative of the composition. In the case of LHST, surfactant components introduced during the washing and shower process, oil components and dispersed particles dissolved by the surfactant accumulate inside the tank, making sludge difficult to remove. In addition, even if it is removed by ultra-high pressure spraying, unexpected problems may occur in the subsequent treatment process due to the surfactant contained therein. Therefore, it is necessary to analyze in more detail the characteristics of sludge accumulated in LHST and prepare countermeasures. A test procedure was prepared to evaluate the characteristics of sludge accumulating in LHST. According to the test results, the long-term sludge accumulation tendency of the LHST is summarized as follows. ① Initially, the sludge settling speed increases slowly until a surface sludge layer is formed. ② After the surface sludge layer is formed, the sludge rapidly settles until the sludge layer becomes somewhat thicker. ③ When the sludge layer is formed to a certain extent, the sludge escape rate increases and the sludge accumulation rate decreases again. It is assumed that the sludge escape speed is closely related to the fluid flow speed in the relevant area. It is believed that the combined effect of these phenomena will determine the thickness of the sludge layer that will accumulate inside the tank, but it was not possible to evaluate how much the sludge layer would accumulate based on the experimental results alone. However, it can be assumed that significant sludge accumulation occurred in areas where fluid flow was minimal and sludge formation nuclei easily accumulates.

According to acceptance of radioactive waste, homogeneous waste such as concentrated liquid waste and spent resin must be solidified to reduce radiological hazards and protect public health and the ecology. However, when using a High Integrity Containers (HIC), it is stated that homogeneous waste can be disposed of without applying the solidification test requirements. PCHIC, developed in korea, is composed of polyethylene (PE, interior), polymer concrete (PC, filler), and steel (external reinforcement). Currently, PC-HIC will be used as a packaging container for low-level liquid waste and spent resin waste. PE has a lower shielding efficiency compared to PC, but it offers the economic advantage of lower production costs. Therefore, cost savings can be expected if very low-level waste is packaged and disposed of HIC made only of PE materials (PEHIC). Despite the economical advantage of PE-HIC, PE-HIC has not been used domestically since NRC (Nuclear Regulatory Commission) reported that PE-HIC couldn’t meet the mechanical integrity criteria for radiation exsure. However, according to IAEA (International Atomic Energy Agency) research, it has been reported that mechanical integrity of PE-HIC is not affected when the absorbed dose is below 50 kGy. Therefore, in this study, Radiological impact of VLLW packaged into PE-HIC is evaluated to confirm that the absorbed dose is below 50 kGy, which then be used to assess feasibility of PE-HIC to be used as packaging and disposal container for radioactive waste. Radiological impact of VLLW packaged into PE-HIC is evaluated to confirm that the absorbed dose is below 50 kGy, which then be used to assess feasibility of PE-HIC to be used as packaging and disposal container for radioactive waste. The feasibility of using PE-HIC as packaging-disposal containers for radioactive waste will be reviewed. In this study, the radiation effects of only waste packaged in PE-HIC were considered, and additional assumptions are as follows. - Nuclides subject to radioactivity evaluation: Co-60, Cs-137 - Radioactivity concentration: very low-level radioactive wastel concentration limit - Target waste: waste resin - PE-HIC dimensions: outer diameter: 1,194 mm, height: 1,290 mm, and thickness 88 mm (PCHIC internal PE shape) Considering the above assumption, the exposure rate was evaluated using the MicroShield program. Since the density of PE-HIC in the MicroShield program was assumed as the density of air. The absorbed dose was recalculated through density correction of the derived exposure rate. As a result, it was confirmed that absorbed dose was about 2-3 mGy over 300 years. As a result of dose evaluation by MicroShield, it is judged that the mechanical integrity of PEHIC as an packaging of VLLW can be proved by confirming that the absorption dose irradiated to PE-HIC by internal waste is less than 50 kGy.

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

Radioactive contamination distribution in nuclear facilities is typically measured and analyzed using radiation sensors. Since generally used detection sensors have relatively high efficiency, it is difficult to apply them to a high radiation field. Therefore, shielding/collimators and small size detectors are typically used. Nevertheless, problems of pulse accumulation and dead time still remain. This can cause measurement errors and distort the energy spectrum. In this study, this problem was confirmed through experiments, and signal pile-up and dead time correction studies were performed. A detection system combining a GAGG sensor and SiPM with a size of 10 mm × 10 mm × 10 mm was used, and GAGG radiation characteristics were evaluated for each radiation dose (0.001~57 mSv/h). As a result, efficiency increased as the dose increased, but the energy spectrum tended to shift to the left. At a radiation dose intensity of 400 Ci (14.8 TBq), a collimator was additionally installed, but efficiency decreased and the spectrum was distorted. It was analyzed that signal loss occurred when more than 1 million particles were incident on the detector. In this high-radioactivity area, quantitative analysis is likely to be difficult due to spectral distortion, and this needs to be supplemented through a correction algorithm. In recent research cases, the development of correction algorithms using MCNP and AI is being actively carried out around the world, and more than 98% of the signals have been corrected and the spectrum has been restored. Nevertheless, the artificial intelligence (AI) results were based on only 2-3 overlapping pulse data and did not consider the effect of noise, so they did not solve realistic problems. Additional research is needed. In the future, we plan to conduct signal correction research using ≈10×10 mm small size detectors (GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the high radiation field of an actual nuclear facility.

In order to evaluate the integrity of the reactor pressure vessel, various test specimens necessary to identify irradiation embrittlement. The degree of irradiation embrittlement of the vessel material by neutrons, from the construction to the end of the life of the plant, is evaluated by a monitoring plan that called surveillance program (a series of all plans to analyze and evaluate embrittlement through various tests and analyzes by placing a test piece inside the reactor pressure vessel and taking out a piece at an appropriate time according to the number of operation years and taking necessary measures for safe operation). The reactor monitoring specimens for Kori Unit-1 are located by axis at S (57°), T (67°), R (77°), N (237°), P (247°) and V (257°). Six surveillance capsules are attached to the inside of the pressure vessel around the core and to the outside of the thermal shield. This surveillance container determines the withdrawal timing of the surveillance container according to the provisions of ASTM E185-82. In the monitoring test piece, there are neutron dosimeter materials to measure and evaluate the irradiated neutron flux, and Ni, Cu, Fe, Co-Al, Cd, and shielded Co-Al monitors are wired in the monitoring container. Each axial position is contained in a spacer hole. The neutron dosimetry monitor measures the neutron dose using isotopes produced by neutrons during operation of the reactor. The Al-Co specimen, which can evaluate the degree of radioactivity of cobalt, is located on the lower part of the specimen. The content of Co in the Al-Co specimen is 0.15%, and when expressed in ppm, it is 1,500 ppm, which is similar to the cobalt content of 1,414 ppm in the internal structure of the reactor vessel presented in NUREG-3474. If the radiation value of the Al-Co sample in the reactor monitoring specimen can be measured, the radiation value of the internal structure of the reactor can be indirectly compared. Since the monitoring specimen is located outside of the thermal shield, radiation should be less than that of the thermal shield. Korea Reactor Monitoring Technology performed gamma measurement on Al-Co specimens in 6 monitoring specimens, and although there are differences depending on the sample, it shows radioactivity values around the order of 1E+07 dps/g, or Bq/g. In conclusion, it is thought that using this measurement values, it is possible to verify the evaluation of internal structure radiation for Kori unit-1 decommissioning.