Safe radiation management is essential not only for operational nuclear power plants but also for nuclear plants to be decommissioned. When spent nuclear fuel is present on-site, meticulous radiation emergency plans are necessary to ensure safety. In Korea, numerous radiation emergency plans have been established for operational nuclear reactors. These plans delineate distinct response mitigation measures for white, blue, and red emergencies. However, clear regulations are yet to be devised for radiation emergency plans for reactors to be decommission. Therefore, this study investigated the decommissioning plan and status of Kori unit 1 to comprehensively analyze the current status of decommissioning safety in Korea. In this study, radiation emergency plans of decommissioning nuclear power plants abroad were reviewed to confirm radiation emergency action levels. Furthermore, radioactive waste treatment facilities, to be used for decommissioning reactors in Korea were evaluated. Moreover, the study assessed emergency plans (especially, emergency initiating conditions) for operational nuclear power plants in Korea for potential use in the decommissioning phase. This study proposed an emergency initiating condition that can be used for decommissioning reactors in Korea. Considering the anticipated introduction of plasma torch melting facility in Korea, this study examined the conditions of radiation emergency plans can be altered. This study identified effective measures and guidelines for managing radiological emergency initiating conditions, and effective decommissioning of nuclear power plants in Korea.

Chelating agents in low and intermediate radioactive wastes can form complexes with radionuclides and increase the mobility of the radionuclides. According to the Korea Radioactive Waste Agency (Acceptance criteria for low and intermediate radioactive waste, WAC-SIL-2022-1), if the amount of residual chelating agents in the waste are greater than 0.1%, the chemical names and residual amounts should be specified; if greater than 1%, the waste must be solidified and contain no more than 8%. The existing method for analyzing chelates in radioactive waste was based on UV–Visible spectrophotometry (UV-Vis), but the new method is based on liquid chromatography/mass spectrometry (LC-MS). The analysis was performed in aqueous solution before applying to real samples. Since the real sample may contain several heavy metals, it is expected that the chelates will exist as complexes. Therefore, 1.0×10-4 mol L-1 of EDTA (Ethylenediaminetetraacetic acid), DTPA (Diethylenetriaminepentaacetic acid), NTA (Nitrilotriacetic acid), and excess metals in aqueous solution were analyzed using HPLC using RP (Reverse Phase) column and HILIC (Hydrophilic interaction) column. When the RP column was used, each substance eluted without separation at the beginning of the analysis. However, when analyzed using a HILIC column, the peaks of each substance were separated. LC-MS measurements using HILIC conditions resulted in separations with better sensitivity.

As existing nuclear power plants reach the end of their lifespan, 22 nuclear power plants in korea are scheduled to be permanently shut down and decommissioned by 2050. Chelates are used as decontamination agents during nuclear power plant operation and decommissioning, and as a result, decommissioning waste contains chelates. Chelates contained in radioactive waste are complexed with radionuclides and increases their mobility. So, qualitative and quantitative analysis of chelates contained in radioactive waste is necessary. However, the spectroscopic method (UVVis), previously used for chelate analysis in Korea takes too much time for analysis and cannot analyze two or more chemically similar chelates at the same time. Due to these problems, new methods for analyzing chelate must be developed. Overseas, many cases of chelate analysis using advanced analysis equipment have been reported. CEA in France has developed a chelate analysis method for application to radioactive waste using HPLC-MS (J. Chromatogram. A, 1276, 20-25, 2013). In this method, the existing method of measuring EDTA using a complex of Fe and EDTA was improved to measuring a complex of Ni and EDTA. Based on such overseas cases, we would like to develop an analysis method for chelates in radioactive waste. For this purpose, we will verify similar overseas papers and develop pretreatment methods for mixtures of chelates (EDTA, DTPA, NTA) and metals (Fe, Ni, Cu, etc.) in various media. Finally, we will develop a separation analysis technology for multi-component chelates in nuclear decommissioning waste based on LCMS.

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.

As nuclear decommissioning ventures become increasingly complex, the role of digitalization in facilitating and enhancing these operations is becoming indispensable. This transition to a more digitized approach presents a myriad of advantages, including: augmented avenues for data acquisition, analysis, and visualization to bolster dismantling strategies; simulations in virtual environments for operator training; precise forecasting of future waste emergence, culminating in refined cost estimations; and more immersive decommissioning visualizations for both operators and external stakeholders. Salient benefits conferred by the integration of digital technologies in decommissioning encompass improved collaboration, enriched knowledge transfer, clarity regarding present technological constraints, insights into key influencing factors, clearer criteria for technology selection, and a profound understanding of the potential challenges and merits of a broader incorporation of digital tools in decommissioning endeavors. Of paramount importance is the opportunity presented for superior workforce training and safety measures, exemplified by ALARAbased planning. Amidst the myriad facets of digital adoption, 3D modeling of nuclear facilities derived from laser-scanned point clouds stands out as a pivotal domain in the digitalization. The transformation of intricate point cloud data into a comprehensible 3D mesh remains the crux of this paper. The process of mesh generation, despite being simpler than its counterpart of converting to a 3D solid model, is crucial for multiple reasons. The resultant 3D mesh offers an enhanced visual representation compared to a sparse point cloud, paving the way for improved spatial perception. Furthermore, it serves as a rudimentary tool for approximating component volumes and the ensuing waste, thereby playing an instrumental role in waste manipulation strategies, notably in collision detection. This paper delves deep into the nuances of mesh generation, conducting an parametric study of mesh conversion algorithms, including down-sampling rates. Through this rigorous examination, we endeavor to shed light on optimal methodologies, hoping to catalyze advancements in the digitalization of nuclear decommissioning processes.

In the dismantling of nuclear power plants, various forms of radioactive gaseous waste are generated when cutting concrete and metal structures. Large amounts of radioactive dust and aerosols generated during the cutting process of each structure can cause radiation exposure to the environment around the workplace and to the radiation exposure in the body of workers. When cutting structures, water is sprayed to reduce the generation of aerosols, so early saturation of the filter is expected due to radioactive aerosols and fine particles containing a large amount of moisture. A mobile air purification device is being developed to a fast and efficient air purifier that can be used for a long time operation to protect workers from radiation exposure in high radiation areas and to minimize the amount of secondary waste generated. In this paper, the direction for a new concept of unit technology that can achieve the development purpose is described.

Kori Unit 1 nuclear power plant is a pressurized water reactor type with an output of 587 Mwe, which was permanently shut down on June 18, 2017. Currently, the final decommissioning plan (FDP) has been submitted and review is in progress. Once the FDP is approved, it is expected that dismantling will begin with the secondary system, and dismantling work on the primary system of Kori Unit 1 will begin after the spent nuclear fuel is taken out. It is expected that the space where the secondary system has been dismantled can be used as a temporary storage place, and the entire dismantling schedule is expected to proceed without delay. The main equipment of the secondary system is large and heavy. The rotating parts is connected to a single axis with a length of about 40 meters, and is complexly installed over three floors, making accessibility very difficult. A large pipe several kilometers long that supplies various fluids to the secondary system is installed hanging from the ceiling using a hanger between the main devices, and the outer diameter of the pipe is wrapped with insulation material to keep warm. In nuclear secondary system decommissioning, it is very important to check for radiation contamination, establish and implement countermeasures, and predict and manage safety and environmental risks that may occur when cutting and dismantling large heavy objects. So we plan to evaluate the radiation contamination characteristics of the secondary system using ISOCS (In- Situ Object Counting System) to check for possible radioactive contamination. According to the characteristics results, decommissioning plans and methods for safe dismantling by workers were studied. In addition, we conducted research on how to safely dismantle the secondary system in terms of industrial safety, such as asbestos, cutting and handling of heavy materials and so on. This study proposes a safe decommissioning method for various risks that may occur when dismantling the secondary system of Kori Unit 1 nuclear power plant.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

In nuclear power plant (NPP) decommissioning, ventilation and purification of the building atmosphere are important to create a working environment, ensure worker safety, and prevent the release of gaseous radioactive materials into the environment. The heating, ventilation, and air conditioning (HVAC) system of each building is maintained, modified, or newly installed. In this study, based on APR1400, operation strategies were presented in case of ventilation abnormalities in the reactor containment building (RCB), where highly radioactive particles and high dust are most frequently generated during NPP decommissioning. For research, it was assumed that the entire RCB atmospheric ventilation during decommissioning would use the RCB purge system of the existing NPP and perform continuous ventilation. Additionally, it is assumed that areas where high radiation particles and high dust occur locally, such as reactor containers or internal segments, are sealed with tents and purified using a HEFA filter of a temporary portable HVAC, and a exhaust flow path is connected to the discharge duct of the existing RCB purge system. The possibility of abnormal occurrence was largely divided into two cases. First, when large amounts of uncontrolled pollutants are released into the atmosphere inside the RCB, discharge to the environment is stopped manually or automatically by a modified engineered safety function activation signal (ESFAS). Afterwards, the RCB purge system should be operated in recirculation mode to sufficiently purify the RCB atmosphere with a HEPA filter. Second, when the first train of the low volume purge system is not running due to a failure, standby train should be operated. If both low volume purge trains fail, a high volume purge system is used. Intermittent purge operation is preferred due to large capacity during high volume purge operation. In cases where it is not possible to operate all purge systems due to common issues such as power supply, atmospheric sampling is performed to determine whether to proceed with the work inside RCB.

Kori unit 1 and Wolsong unit 1 were permanently shut down in 2017 and 2019, respectively. Both plants were decided to demolish the building without reuse. Large structures must be demolished after removing systems and components in the building, and in the case of large structures, thorough planning is required because of the large scale of work. Therefore, in this study, important considerations in the phase of the demolition plan of large structures when decommissioning were analyzed. The demolition of large structures at nuclear facilities is major one phase of work within a broader decommissioning plan. Furthermore, the actual demolition of the structure (i.e., physical process) represents the last step in a process that begins with extensive planning and analysis. The National Demolition Association (NDA) has provided checklist items that should be considered before the start of a commercial demolition project and/or in the bid process. Important Considerations in the Phase of the demolition plan of large structures when decommissioning of nuclear facilities are Site knowledge and programs, Engineering survey/demolition plan, Hazardous and radioactive materials, Open air demolition, Financial and project management, Permits, Code adherence, and Special programs, Disposal pathway, Final site condition. The results of this study can be used as a basis for the Planning large structures demolition of the Kori unit 1 and Wolsong unit 1.

The Korea Research Institute of Standards and Science has developed certified reference materials (concrete, soil, and metal radioactive liquid) for measuring gamma-emitting radionuclides to improve and maintain the quality assurance and quality control of the radioactivity measurement in decommissioning nuclear power plants. The raw materials that make up each CRM were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on the radionuclides in the CRMs fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. In concrete analysis, the reference values for five radionuclides (60Co, 241Am, 134Cs, and 137Cs) in the CRM were in the range of 15-40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In soil analysis, the reference values for the 137Cs and 60Co were 118.7 and 124.4 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In metal radioactive liquid analysis, the reference values for 134Cs, 137Cs and 60Co in the CRM were in the range of 200-270 Bq/kg, and the expanded uncertainty was within 7% (k = 2).

The decommissioning of domestic Nuclear Power Plants (NPPs) in Korea is expected to begin with the Kori-1, which was permanently shutdown in 2017. In addition, Wolsong-1 has been also permanently shutdown, and another type will be the decommissioning project following Kori-1. KHNP is promoting operation and decommissioning projects as the owner of NPPs, and the Central Research Institute (CRI) has been developing a Final Decommissioning Plan (FDP) for the decommissioning license document. The FDP consists of 11 major chapters in the order of overview of the project, characteristic evaluation, safety assessment, radiation protection, decontamination & dismantlement activities, waste management, etc. The contents described in each chapter are individual chapters, but there are also parts that consider the connection with other chapters. The CRI, which develops the FDP for the first decommissioning project in Korea, has spent a lot of time and effort considering this and has been proceeding through trial and error until the present stage. Therefore, this study aims to explain the current status of FDP, a license document for domestic decommissioning projects, and the link between major input data in major chapters. It can be said that System, Structure, and Components (SSCs) subject to dismantling are considered as the scope of FDP. Chapters that perform estimations on these dismantling targets may include safety assessments, exposure dose assessments for workers and residents, and waste inventory assessments. Therefore, an important part of performing the estimation works is to consider the entire scope of decommissioning activities, and as a way, it can start from data based on the inventory data. After generating the inventory data, the waste treatment classification for the inventory is designated by reflecting the results of the characterization. In addition, for cost estimation, the cost of decommissioning project is predicted by inputting some data (i.e., UCF) such as work process, number of workers, and time required for each item with data reflected in quantity and characterization. After that, based on these inventory, characterization, and UCF data, accident scenarios and industrial safety evaluation are performed for the safety assessment. The worker exposure dose is estimated by considering the dose rate of the workspace with these data. In the case of the amount of waste, the final amount of waste is estimated by considering the factors of reduction and decontamination. In summary, the main estimation contents of FDP are evaluated by adding elements required for the purpose of each chapter from data combined with inventory, characterization, and UCF, so the contents of these chapters are based on the logic of considering the entire scope of decommissioning in common.

The Radiation and Decommissioning Laboratory of Central Research Institute (CRI) of Korea Hydro and Nuclear Power Co. (KHNP) performs research to technically support the effective management of radiological hazards to avoid risks to civilians, the workers, and the environment from the radiological risks. The laboratory mainly consists of three technical groups: decommissioning and SF technology group, radiation and chemistry group, and radwaste and environment group. The groups carry out various R&D such as decommissioning, spent fuel management, radiation protection, water chemistry management, and radioactive waste management. The laboratory also technically supports the calibration of radiometric instruments as a Korea Laboratory Accreditation Scheme (KOLAS), approval for decommissioning, guidance for radioactive waste management, state-of-the-art technology evaluations, and technology transfer.

The effects of an individual effective dose from radioactive contamination that will remain during site reuse after the decommissioning of nuclear facilities is generally assessed using the RESRAD code. The calculated results should meet the site reuse criteria presented by regulators, 0.25 mSv/yr in the United States and 0.1 mSv/yr in Korea. After completion of decommissioning, the dose is not subject to measurement, resulting in Derived Concentration Guideline Level (DCGL) remaining at the site that is practically consistent with the dose criteria. In order to assess dose using the RESRAD code, various requirements will need to be considered and determined, where the selection of input parameters is one of the important factors in the dose assessment. In addition, appropriate selection of site-specific parameters is important to reflect the site characteristics of each decommissioned Nuclear Power Plant (NPP). Therefore, this study intends to analyze the impact of site-specific parameters by referring to the cases of overseas decommissioned NPPs. In order to evaluate doses using RESRAD code, a site reuse scenario must first be selected. In general, in the case of unrestricted reuse, the resident farmer scenario can be applied, so the resident farmer scenario was also selected in this study. In addition, once a resident farmer scenario is selected, input parameters are selected according to the scenario, and the input parameter inputs a single value or distribution according to the deterministic or probabilistic evaluation method. Therefore, since this study is to evaluate the effect on site-specific parameters, a single value was applied as a deterministic evaluation method. For the 10 site-specific parameters considered in overseas cases, the difference was set twice using the F9 function key in the RESRAD code and the results were analyzed. In this study, we used prior research data targeting domestic nuclear facility for sensitivity analysis. Related parameters include the category of contamination layer, soil, water transport, ingestion, and occupancy. The parameters that appeared as the greatest influence among the 10 parameters were different in radionuclide on the contaminated zone. We showed the changes according to the difference in input parameters was presented using the graph provided by the RESRAD code. As a result, in the evaluation for Co-60 in this study, no significant change was observed. However, in case of H-3, several parameters values were changed, indicating that the effect on dose will be different depending on the site characteristics of the nuclear facilities.

Safety-related items in the decommissioning Nuclear Power Plants (NPPs) can largely consider safety for workers and residents. At this time, the effects of radioactive contamination on the Systems, Structures, and Components (SSCs) are caused by the performance of work related to Decontamination and Dismantlement (D&D) activities. Classification according to dismantling activities will be important, and the decay factor of radionuclides and the impact of contaminations due to plant characteristic (thermal and electrical capacity) in estimation of exposure dose from such activities will be considered compared to other overseas NPPs. Therefore, this study will consider some factors to consider for comparison with overseas cases in estimating worker exposure dose. To assess worker exposure doses, the classification of decommissioning activities must first be made. It should be classified including large components that can be generally considered, and the contents should be similar to compare with overseas cases. In case of decommissioned NPPs with prior experience, it is possible to predict worker’s exposure with respect to plant capacity, but this does not seem to have a specific correlation when reviewing the related data. Depending on the plant capacity, the occurrence of contamination of radioactive materials may have some correlation, but it cannot be determined that it has causality with the worker’s dose when dismantling. In addition, it is expected that the effects of workers’ exposure doses will vary depending on when the highly contaminated SSCs will be dismantled from permanent shut down. Therefore, the decay correlation coefficient for this high radiation dose works should be considered. If the high radiation dose work is performed before the base year, a correlation coefficient larger than 1 value will be applied, and in the opposite case, a value less than 1 will be applied. Whether or not to perform Full System Decontamination (FSD) is also an important consideration that affects worker dose, and correlation factors should be applied. In this study, the matters to be considered when estimating worker dose for dismantling NPPs were reviewed. This suggests factors to be reflected in the work classification and dose results for comparison with overseas NPP experiences. Therefore, when doing the workers’ dose estimation, it is necessary to derive a normalized doses considering each correlation factor when comparing with overseas cases along with dose estimation for the dismantling activities.

One aspect of securing safety from the operation of Nuclear Power Plants (NPPs) is to evaluate the impact on residents at the facility’s exclusive area boundary to confirm that the radiological risk is below the allowable level. Normally, the risks from gaseous and liquid effluents are evaluated during the operation of facilities. Meanwhile, in order to be approved for the decommissioning plan, the environmental risks caused by activities during dismantling is also evaluated. Therefore, this study aims to investigate the exposure pathways considered in evaluating the risks to nearby residents from the operation and decommissioning of nuclear facilities and to examine the differences. The emission rate by radionuclide is calculated by evaluating the amount of leak from nuclear fuel during the operation of the facility through design data of the NPP. Each of the liquid and gaseous effluents is calculated, and the exposure dose received by nearby residents is calculated by considering the exposure pathways with these emission rates. In order to initiate the decommissioning of nuclear facilities, approval of the Final Decommissioning Plan (FDP) must be obtained. The FDP chapter shall describe the results of the environmental impact assessment of the decommissioning. It will not differ significantly in the exposure pathways during operation. However, the decommissioning of nuclear facilities is ultimately to remove Systems, Structures, and Components (SSCs) and to remove the regulation of the Nuclear Safety Act by ensuring that sites and remaining buildings meet the criteria for the license termination. In terms of release and reuse of nuclear facilities, the exposure dose to be considered in evaluating the dose can be considered for two main types: the site and the remaining building. The factors affecting the exposure pathways considered in assessing the environmental impacts considered in the operation and decommissioning of nuclear facilities are due to gaseous and liquid effluents. However, the difference should reflect the impact of NPP operations and decommissioning activities when evaluating the amount of radionuclides released by these effluents. Decommissioning should consider the impact after decommissioning, which is the effect of the receptor by radionuclides remaining on the site and in the remaining buildings. At this time, the effects of the source from the soil and the source from the surface of the building should be considered for the external and internal exposure pathways.

The decommissioning of the Nuclear Power Plant (NPP) is a long-term project of more than 15 years and will be carried out as a project, which will require project management skills accordingly. The risk of decommissioning project is a combination of many factors such as the decommissioning plan, the matters licensed by the regulatory agency, the design and implementation of dismantling, the dismantling plan and organization, and stakeholders. There will be some difficulties in risk management because key assumptions about many factors and the contents of major risks should be well considered. Risk management typically performs a series of processes ranging from identification and analysis to evaluation. In order to analyze and evaluate risks here, identification of potential risks is the first step, and in order to reasonably select potential risks, various factors mentioned should be considered. Therefore, the purpose of this study is to identify possible risks that should be considered for the decommissioning project in various aspects. The risk of the decommissioning project can be defined using the hazard keyword, and the risk family presented in the IAEA safety series can also be referred. It would be better to approach the radiological or non-radiological risks that may occur in the dismantling work with the hazard keyword, and if the characteristics of the decommissioning project are reflected, it would be a good idea to approach it on a risk family basis. There are 10 top risks in the risk family, 25 risks at the level 2 and 61 risks at the level 3 are presented. It may be complex to consider these hazards and risks recommended as risk families at the same time, so using the results of safety evaluation as input data for risk identification can be a reasonable approach. Therefore, this study intended to derive the possible risks of the decommissioning project based on the risk family structure. At this point, the reflection of the safety assessment results was intended to be materialized by considering the hazards checklist. As a result, this study defined and example of 38 possible risks for the decommissioning project, considering the 10 top risk family and lower level risk categories. This result is not finalized, and it will be necessary to further strengthened through expert workshops or HAZOP in the future.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

After permanent shutdown, contamination existing in nuclear facilities must be removed according to decontamination and dismantling procedures to achieve the target end state. In Korea, Korea Research Reactor (KRR) Units 1, 2 are being decommissioned, and Kori Unit 1 is in the process of reviewing the final decommissioning plan for the start of decommissioning. In order to complete decommissioning of nuclear facilities, it is necessary to satisfy the dose criteria according to the residual radioactivity remaining in the site and buildings. In the United States, which has a lot of experience in decommissioning, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) is used as a procedure for measuring and analyzing residual radioactivity. In MARSSIM, survey units are classified according to the level of contamination, and the radiation survey procedure and effort can be determined according to the survey unit level. After the radiological analysis and statistical verification of the survey unit, it is decided whether to release the site. At this time, the geographical area to be used as the background level is called the reference area. Therefore, selection of an appropriate reference area is important for accurate residual radioactivity analysis and for the release of the site. In this study, reference area evaluation cases and domestic decommissioning procedures were analyzed to derive considerations for selecting an appropriate reference area. For example, Zion NPP in the US selected a place outside the boundary of the restricted area unaffected by nuclear power plant operation as a reference area by referring to the meteorological monitoring report. Among Korea’s decommissioning procedures, the appropriateness of the reference area can be confirmed through the final status report submitted upon completion of decommissioning. However, since the selection and application of the reference area needs to be reflected during decommissioning, relevant information must be updated through periodic communication between operator and regulatory agency. The results of this study will be used as considerations for selecting a reference area.

To evaluate the characteristics of radioactive waste from permanently shut down nuclear power plants for decommissioning, there is a method of directly analyzing samples and, on the other hand, a computerized evaluation method based on operation history. Even if the radioactivity of the structures or radioactive wastes in the nuclear power plant is analyzed by the computerized evaluation method, a method of directly analyzing the sample must be accompanied in order to more accurately know the characteristics of the nuclear power plant’s radioactive waste material. In order to obtain such samples, we need a way to collect materials from radioactive waste. However, in the case of a permanently shut down nuclear power plant with a long operating history, human access is limited due to radiation of the material. In this study, we propose a method of remotely collecting samples that guarantees radiation protection and worker safety at the site where radioactive waste is located.