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.

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.

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.

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.

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.

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.

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.

Currently, Korea is planning to use various equipments and technologies for cutting, decontamination, compression, solidification, and packaging at decommissioning site of Kori unit 1 and Wolsung unit 1. In particular, Korea is considering to apply new technologies like inorganic acid decontamination, spent resion treatment technology not only to localize various decommissioning technology, but to meet the limit of 14,500 drums of the decommissioning waste per unit. However, before the techniques applied to decommissioning, it is necessary to demonstrate the effectiveness and the safety of the techniques. Because unlike the industrial fields, the failure of the decommissioning technique in nuclear power plants can cause the spread or leakage of radioactive materials. In the「Regulation on Technical Standards for Nuclear Reactor Facilities, Etc.」is stated that the licensee shall apply proven technology to decommissioning and if the licensee apply new technology, he must provide resonable evidence and prove its safety. In accordance with this approach, Nuclear Safety and Security Commission (NSSC) Notice No. 2021-24 can be applied to the decommissioning technology as one of a technical standard related with the demonstration of it. And it states 9 kinds of elements related to the radioactive waste management facilites and components like the management of radioactive effluents, prevention of contamination and overflow by radioactive materials, etc. But, they are mixed with the radioactive material considerations and industrial considerations, and these considerations are usually for the facilities, not equipments or techniques. On the other hand, in the IAEA Safety Standards Series No. WS-G-2.1 Section 6.15 to 6.20, it recommended to evaluate 12 considerations for the decontamination technique and 7 considerations for the dismantling techniques. The Decommissioning Guide in Germany recommends to consider 3 conditions for radiation protection of decontamination techniques and 4 conditions for dismantling techniques. Therefore, it is necessary to compare the safety requirements or recommendations related to the demonstration of decommissioning technology with the other countries to check there is something to learn from it.

The nuclear power plant (NPP) decommissioning market is expected to expand not only domestically but also overseas. Proven technologies must be applied to decommission NPP. This is based on Article 41-2, Paragraph 2 of the domestic ‘Enforcement Decree Of The Nuclear Safety Act’. Proven technology refers to technology that has verified that it can be applied in the field through demonstration. In other words, in order to carry out NPP decommissioning, verification must be done. Demonstration refers to reducing technological uncertainty and directly verifying services implemented in the field. From a technology commercialization perspective, demonstration requires an approach based on technology readiness level (TRL) from a technology perspective and market readiness level (MRL) from a market perspective. The characteristics of demonstration also differ depending on the characteristics of each field. The demonstration in the field of nuclear energy is the demonstration of demand matching. This is to confirm the feasibility of the technology in the company’s required environment. In order to perform demonstration, a scenario must be derived by reflecting demonstration design considerations. After evaluating the derived scenario, an actual assessment is conducted using lab-based demonstration/virtual environment demonstration/real environment demonstration. What must be preceded by an actual assessment is confirming the consumer’s requirements. In this study, the necessary environment and requirements of consumer’s to perform NPP decommissioning were reviewed. The domestic decommissioning procedure requirements management system presents decommissioning procedures, potential worker accidents, and worker requirements. In the case of foreign countries, it was confirmed that complex wide need, cost benefit, risk reduction, waste generation, operation, reliability and maintenance (RAM) improvement and quantitative measures were evaluated for the technology to be demonstrated. Also the requirements for demonstrating decommissioning need to a detailed review of actual decommissioning cases. Therefore, a comparison must be made between the requirements based on actual NPP decommissioning cases and the requirements derived from this research process. Afterwards, the empirical research approach proposed by the Ministry of Trade, Industry and Energy was applied. The empirical research approach proposed by the Ministry of Trade, Industry and Energy is to secure a track record over a certain period of time and performance under conditions similar to the actual environment in the final research stage at the TRL level 6 to 8. Through this, it will be possible to confirm the suitability of overseas technology for domestic application.

KHNP is conducting research to decommission Wolsong Unit 1 Calandria. Establishment of preparation and dismantlement processes, conceptual design of equipment and temporary radiation protection facilities, and waste management are being established. In particular, the ALARA plan is to be established by performing exposure dose evaluation for workers. This study aims to deal with the methodology of evaluating exposure dose based on the calandria dismantling process. The preparation process consists of bringing in and installing tooling and devices, and removing interference facilities to secure work space. The main source term for the preparation process is the calandria structure itself and crud of feeders. In the case of the dismantlement process, a structure with a shape that changes according to the process was modeled as a radiation source. It is intended to estimate the exposure dose by selecting the number of workers, time, and location required for each process in the radiation field evaluated according to the preparation and dismantlement process. In addition, it is also conducting an evaluation of the impact on dust generated by cutting operations and the human impact of C-14, H-3, which are specialized nuclides for heavy water reactors. KHNP is conducting an exposure dose evaluation based on a process based on the preparation and dismantlement process for decommissioning Calandria through computation code analysis. If additional worker protection measures are deemed necessary through dose evaluation according to this methodology, the process is improved to prepare for the dismantling of worker safety priorities.

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.

Derived Concentration Guideline Levels (DCGLs), which represent the residual radioactivity concentration limits, serve as the pivotal criteria for decontamination during decommissioning of nuclear power plants and are essential for license termination. The analysis of radionuclides in various media to check site-specific and radionuclide-specific DCGLs is a resource-intensive and time-consuming processes, and there are some radionuclides that are hard to analyze. In the decommissioning of the Rancho Seco nuclear power plant in the United States, a conservative approach was adopted. Potentially highly contaminated areas on the site were identified by collecting and analyzing soil samples, and radionuclides exceeding the Minimum Detectable Concentration (MDC) were selected as the potential Radionuclide of Concern (ROC), and surrogate DCGLs for hard-to-detect radionuclides were applied to soil samples. For soil samples in the Rancho Seco nuclear power plant, Cs-137 contributed more than 90% of the total radioactivity. DCGLs of the ROC were obtained using the scaling factors through analysis of Cs- 137 for a large amount of soil samples. In Korea, the scaling factor methodology has not been applied to the decommissioning of commercial nuclear power plants. An initial investigation was undertaken to assess the viability of implementing Surrogate Derived Concentration Guideline Levels (DCGLs) in the dismantling of Kori Unit 1, drawing insights from the U.S. nuclear power plant decommissioning experiences. To do this approach, the concentration ratio of radionuclides of interest to key radionuclide in contaminated soil should be known and consistent. But related information is not available at this time. So Surrogate DCGL for representative C-14, Fe-55, Ni-59, Ni-63, and Sr-90 was obtained using the scaling factors applied to radioactive waste data, specifically Decontaminated Aqueous Waste (DAW) and Spent Resin. In order to develop a reliable surrogate DCGLs the Kori Unit 1 site, it is important to analyze the radionuclides in the soil for the Kori Unit 1 decommissioning site to obtain consistent concentration ratio of the radionuclides of concern to the key radionuclides. When a the suitable DCGL is developed, it can be used for FSS planning and prior decision-making ensuring the safe and effective decommissioning of Kori Unit 1 and similar nuclear power plants.

The domestic Pressurized Heavy Water Reactor (PWHR) nuclear power plant, Wolsong Unit 1, was permanently shut down on December 24, 2019. However, research on decommissioning has mainly focused on Pressurized Water Reactors (PWRs), with a notable absence of both domestic and international experience in the decommissioning of PHWRs. If proper business management such as radiation safety and waste is not performed, it can lead to increased business risks and costs in decommissioning. Therefore, the assessment of waste volume and cost, which provide fundamental data for the nuclear decommissioning process, is a crucial technical requirement before initiating the actual decommissioning of Wolsong Unit 1. Decommissioning radiation-contaminated structures and facilities presents significant challenges due to high radiation levels, making it difficult for workers to access these areas. Therefore, technology development should precede decommissioning process assessments and safety evaluations, facilitating the derivation of optimal decommissioning procedures and ensuring worker safety while enhancing the efficiency of decommissioning operations. In this study, we have developed a program to estimate decommissioning waste amounts for PHWRs, building upon prior research on PWR decommissioning projects while accounting for the specific design characteristics of PHWRs. To evaluate the amount of radioactive waste generated during decommissioning, we considered the characteristics of radioactive waste, disposal methods, packaging container specifications, and the criteria for the transfer of radioactive waste to disposal operators. Based on the derived algorithm, we conducted a detailed design and implemented the program. The proposed program is based on 3D modeling of the decommissioning components and the calculation of the Work Difficulty Factor (WDF), which is used to determine the time weighting factors for each task. Program users can select the cutting and packaging conditions for decommissioning components, estimate waste amount based on the chosen decommissioning method, and calculate costs using time weighting factors. It can be applied not only to PHWRs, but also to PWRs and non-nuclear fields, providing a flexible tool for optimizing decommissioning process.

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.

Carbon 14 (14C) is radioactive isotope of carbon which emits beta ray with long half-life (5730±30 years). Since the 14C is significantly hazardous for human being, the appropriate process to treat 14C is necessary. From the nuclear power plant, the ion exchange resin, graphite, and activated carbon are the main source of 14C. During the effort to reduce the volume of those wastes, the 14C is inevitably occurred as carbon dioxide (CO2) form, so called 14CO2. Thus, the development of technology to permanently capture and safely dispose 14CO2 is required. In this presentation, we introduce the decommissioning technology ranging from 14CO2 capture to solidification. First, the new class of glass adsorbent is developed which can irreversibly capture CO2 even under mild conditions. This material promotes the dissolution of alkaline earth ions due to the unstable glass structure. Then, the physical and chemical optimization of glass adsorbent enhances the performance of CO2 capture. Further, room temperature geopolymeric solidification is also performed to safely dispose 14C without any potential release.

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.