Protoplasts were isolated from the primary leaves of lettuce (Lactuca sativa L.) seedlings 10 days after in vitro germination. The leaves were stripped and incubated in an enzyme mixture consisting of 1.2% Cellulase R-10 and 0.3% Macerozyme R-10 in cell and protoplast washing solution (CPW) overnight. The average protoplast yield was 8.25 x 106 protoplasts per g of fresh leaf tissue. When protoplasts were cultured at a density of 3.0 × 105 protoplasts/mL in agarose solid KM8P/KM8 medium, first and second divisions were observed in the protoplasts within a week. Protoplast-derived microcolonies formed after 4 weeks of culture, and visible colonies were present after 3 months of culture. Protoplast-derived microcalli were transferred to Murashige and Skoog medium supplemented with 2.0 mg/L kinetin and 0.1 mg/L NAA and incubated in the light for 3 weeks. They grew into callus, which then regenerated into plants after 7 weeks of culture. The regenerated plants grew as apparently normal flowering fertile plants.

The current study, which consisted of two independent studies (laboratory and greenhouse), was carried out to project the hypothesis fungi-spray scheduling for leaf mold and gray leaf spot in tomato, as well as to evaluate the effect of temperature and leaf wet duration on the effectiveness of different fungicides against these diseases. In the first experiment, tomato leaves were infected with 1 × 104 conidia·mL-1 and put in a dew chamber for 0 to 18 hours at 10 to 25°C (Fulvia fulva) and 10 to 30°C (Stemphylium lycopersici). In farm study, tomato plants were treated for 240 hours with diluted (1,000 times) 30% trimidazole, 50% polyoxin B, and 40% iminoctadine tris (Belkut) for protection of leaf mold, and 10% etridiazole + 55% thiophanate-methyl (Gajiran), and 15% tribasic copper sulfate (Sebinna) for protection of gray leaf spot. In laboratory test, leaf condensation on the leaves of tomato plants were emerged after 9 hrs. of incubation. In conclusion, the incidence degree of leaf mold and gray leaf spot disease on tomato plants shows that it is very closely related to formation of leaf condensation, therefore the incidence of leaf mold was greater at 20 and 15°C, while 25 and 20°C enhanced the incidence of gray leaf spot. The incidence of leaf mold and gray leaf spot developed 20 days after inoculation, and the latency period was estimated to be 14‒15 days. Trihumin fungicide had the maximum effectiveness up to 168 hours of fungicides at 12 hours of wet duration in leaf mold, whereas Gajiran fungicide had the highest control (93%) against gray leaf spot up to 144 hours. All the chemicals showed an around 30‒50% decrease in effectiveness after 240 hours of treatment. The model predictions in present study could be help in timely, effective and ecofriendly management of leaf mold disease in tomato.

담낭암은 드물지만 치명적인 악성 종양으로 초기에는 증상이 없어 대부분의 담낭암이 늦게 진단되고, 급속하게 인접 장기로 전이되기 때문에 예후가 불량하다. 따라서 담낭암을 조기에 발견하는 경우는 드물고, 담낭담석이나 용종으로 수술적 절제를 시행한 후 우연히 발견되는 때가 대부분이다. 저자들은 급성 무결석 담낭염 환자에서 초음파 내시경을 통해 비교적 조기에 담낭관암을 발견하여 성공적으로 치료할 수 있었던 증례를 경험하였다. 이에 문헌고찰과 함께 보고한다.

With respect to the geologic repository, intrusion of groundwater has been considered as a major factor that can transfer radionuclides to the natural environment. Moreover, the migration of radionuclides in the natural groundwater system is significantly influenced by the interaction between the radionuclides and groundwater constituents. Among various hydrogeochemical reactions, hydrolysis is one of the major reactions that can affect the aqueous solubility of radionuclides. Therefore, a precise understanding of relevant chemical thermodynamic behavior is of cardinal importance for the reliable prediction of migration/retardation behavior of radionuclides in the natural groundwater system. The objective of the present work is to investigate the solubility behavior of Nd(OH)3(s) to provide relevant chemical thermodynamic data of Nd(III) as a chemical analogy of major radiotoxic elements such as Am(III) and Cm(III). All the experiments were performed with Ar gas-filled glovebox under inert atmospheric condition. The aqueous Nd(III) solution was prepared by dissolution of 0.5 g NdCl3·6H2O (Sigma-Aldrich) in 10 ml of deionized water. The Nd(III) solid phase was precipitated by dropwise addition of ca. 10 ml of 4 M NaOH (Sigma-Aldrich). The Nd(III) precipitate was identified to be crystalline Nd(OH)3(s) nanorod by using XRD and TEM. For the solubility experiment, the solid Nd(OH)3(s) was equilibrated at the pH range from 5.0 to 9.0 at 0.1 M NaCl condition. The total concentration of the Nd(III) was quantified by using UV/Vis absorption spectroscopy and ICP-MS after the phase separation. In the present work, the solubility behavior of the solid Nd(OH)3(s) phase was investigated by using colorimetric analysis. The chemical thermodynamic data obtained in this study are expected to enhance the reliability of solubility prediction for the trivalent lanthanides and actinides.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

The Korea Atomic Energy Research Institute operates the Nuclear Cycle Experimental Research Facility which has radiation controlled area in the laboratory with the aim of realizing pyroprocessing technology. In this Facility, depleted Uranium feed material and a depleted Uranium mixed with some surrogate material are used for performing experiments. Therefore the facility is using uranium, users should be careful of radiation. This paper will explain the radiation protection of the Nuclear Cycle Experimental Research facility and will also explain how much alpha radiation comes out from the facility. The RMS (Radiation Monitoring System) detector is made by CANBERRA and the model name is ICAM. ICAM RMS is the detector which can detect Alpha Radiation by absorbing the air in the facility. The RMS detector is installed in the HVAC room on the third floor to check the air contamination through the chimney. The RMS is connected to the air ventilation line for detecting Alpha radiation in the whole facility. Experiment was performed for two weeks to check the radiation level and the air ventilation fan continued to operate 24 hours a day. the results are below the required value which is 0.1 Bq/m3, indicating that the facility is safe in terms of radiation safety management.

The dimensioning machine installed in the hot cell has been used for 20 years. It has been used for a long time so it was often malfunction due to aging and radiation. In addition, some parts of apparatus were discontinued and there were a lot of problems in maintenance and repair. In the old measuring system, the operator’s subjectivity was much involved. The process of control the focal length (distance between lens and specimen) by operator’s sense is a good example. The existing dimensioning machine was the Kh-7700 of Hirox Co., Ltd. As the equipment had been used for a longtime, additional utilities such as jigs, lighting module and servo motors have been customized and used. The same company’s apparatus was selected for the reasons that it did not need to manufacture a new utility so it could save the cost of radioactive waste disposal for existing utilities and its radiation resistance which has been used for 20 years in radiation environment. Lighting, standing, stage, controllers, cables and so on had been customized to provide remote control in hot cell. The installation was completed in March of this year in hot cell and has been successfully used until now. Through the improvement of dimensioning machine, an auto-focusing and multi-focusing were available. Therefore, they made it possible to produce high quality data and improve the accuracy of data. And this research could be a good example of how hot cell devices can be built and customized to achieve remote control.

As the decommissioning and decontamination (D&D) of nuclear power plants (NPPs) has actively proceeded worldwide, the management of radiation exposure of workers has become more critical. Radioactive aerosol is one of the main causes of worker exposure, contributing to internal exposure by inhalation. It occurs in the process of cutting radioactive metal structures or melting radioactive wastes during D&D, and its distribution varies according to decommissioning strategies and cutting methods. Among the dominant radionuclides in radioactive aerosols, Fe-55 is known to be the most abundant. Fe-55, which decays by electron capture, is classified as a difficult-to-measure (DTM) radionuclide because its emitted X-rays have too low energy to measure directly from outside of the container. Generally, for measuring DTM nuclides, the liquid scintillation counting (LSC) method and the scaling factor (SF) method are used. However, these methods are not suitable for continuous monitoring of the D&D workplace due to the necessity of sampling and additional analysis. The radiation measurement system that can directly measure the radionuclides collected at the aerosol filter could be more useful. In this study, as preliminary research on developing the radioactive aerosol monitoring system, we fabricated a gamma-ray spectrometer based on a NaI (Tl) scintillator and measured the energy spectrum of Fe-55. A beryllium window was applied to the scintillator for X-ray transmission, and the Fe-55 check source was directly attached to the scintillator assuming that the aerosol filter was equipped. 5.9 keV photopeak was clearly observed and the energy resolution was estimated as 44.10%. Also, the simultaneous measurement with Cs-137 was carried out and all the peaks were measured.

Plastic scintillators can be used to find radioactive sources for portal monitoring due to their advantages such as faster decay time, non-hygroscopicity, relatively low manufacturing cost, robustness, and easy processing. However, plastic scintillators have too low density and effective atomic number, and they are not appropriate to be used to identify radionuclides directly. In this study, we devise the radiation sensor using a plastic scintillator with holes filled with bismuth nanoparticles to make up for the limitations of plastic materials. We use MCNP (Monte Carlo N-particle) simulating program to confirm the performance of bismuth nanoparticles in the plastic scintillators. The photoelectric peak is found in the bismuth-loaded plastic scintillator by subtracting the energy spectrum from that of the standard plastic scintillator. The height and diameter of the simulated plastic scintillator are 3 and 5 cm, respectively, and it has 19 holes whose depth and diameter are 2.5 and 0.2 cm, respectively. As a gamma-ray source, Cs-137 which emits 662 keV energy is used. The clear energy peak is observed in the subtracted spectrum, the full width at half maximum (FWHM) and the energy resolution are calculated to evaluate the performance of the proposed radiation sensor. The FWHM of the peak and the energy resolution are 61.18 keV and 9.242% at 662 keV, respectively.

Gamma spectrometry is one of the main analysis methods used to obtain information about unknown radioactive materials. In gamma-ray energy spectrometry, even for the same gamma-ray spectrum, the analysis results may be slightly different depending on the skill of the analyst. Therefore, it is important to increase the proficiency of the analyst in order to derive accurate analysis results. This paper describes the development of the virtual spectrum simulator program for gamma spectrometry training. This simulator program consists of an instructor module and trainee module program based on an integrated server, in which the instructor transmits a virtual spectrum of arbitrarily specified measurement conditions to the students, allowing each student to submit analysis results. It can reproduce a virtual gamma-ray energy spectrum based on virtual reality and augmented reality technique and includes analysis function for the spectrum, allowing users to experience realistic measurement and analysis online. The virtual gamma-ray energy spectrum DB program manages a database including theoretical data obtained by Monte Carlo simulation and actual measured data, which are the basis for creating a virtual spectrum. The currently developed database contains data on HPGe laboratory measurement as well as in-situ measurements (ground surface, decommissioned facility wall, radiowaste drum) of portable HPGe detectors, LaBr3(Ce) detector and NaI detector. The analysis function can be applied not only to the virtual spectrum, but also to the input measured spectrum. The parameters of the peak analysis algorithm are customizable so that even low-resolution spectra can be properly analyzed. The validity of the database and analysis algorithm was verified by comparing with the results derived by the existing analysis programs. In the future, the application of various in-situ gamma spectrometers will be implemented to improve the profiling of the depth distribution of deposited nuclides through dose rate assessment, and the applicability of the completed simulator in actual in-situ gamma spectrometry will be verified.

Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

The IAEA recommended considerations for exemption regulations of consumer products containing greater amounts of radioactive isotopes than the amounts specified for generic exemption. One of the major considerations is the expected exposure dose should be less than 10 μSv/y and 1 mSv/y for general cases and low probability cases, respectively, in all predictable scenarios. Under this recommendation, many countries evaluated the radiation dose for exposure scenarios of various products in consideration of the national circumstances and, then, established their own specific exemption regulation. In Republic of Korea, the “Regulation on substances excluded from radioactive isotopes” was legislated to specify consumer products excluded from regulation. However, as the usage status and product specifications has changed over time, it is necessary to periodically verify the validity of the regulation criteria in the view of exemption justification. In this study, we developed the use and disposal scenarios in consideration of the domestic use of thorium-containing gas mantle and evaluated radiation dose of each scenario accordingly. The gas mantles are used as a wick for gas lanterns and the maximum activity of natural thorium contained among the currently available gas mantles is 12.5 kBq. Radioactive isotopes in the decay chain of natural thorium can be divided into three groups according to their physical characteristics, and exposure routes suitable for each group were considered in dose calculation. Currently, most gas mantles are installed in camping lanterns. Therefore, we developed use scenarios related to camping. The average number of camping trips and time spent at the campground were set by the data from Korea Tourism Organization. Tent sizes and vehicle specifications were determined by referring to surveys and products in Korea. The used gas mantle is disposed of in a garbage bag for general waste and transported to landfill or incinerator. We determined the amount of gas mantle discarded in landfill and incinerator by the data from Korea Environment Corporation. The exposure time and amount handled by an individual were determined by considering the number of waste collection vehicles, landfills, and incinerators. Although we assumed the maximum activity of the gas mantle for conservative evaluation, the calculated radiation doses for the use and disposal scenarios were below the general requirement (i.e., 10 μSv/y) in all scenarios.

Radioactive source terms are important factor in design, licensing and operation of SMR (Small Modular Reactor). In this study, regulatory requirements and evaluation methodology for normal operation on NuScale SMR, which received standard design certification approval on September 11, 2020 from US NRC, are reviewed. The radioactive waste management system of nuclear power reactor should be designed to limit radionuclide concentration in effluents and keep radioactive effluents at restricted area boundary ALARA according to 10 CFR 20 and 10 CFR 50 Appendix I. Also, in general, the coolant source term to calculate the off-site radiological consequences for normal operation of SMR should be determined by using models and parameters that are consistent with regulatory guide 1.112, NUREG- 0017 and the guidance provided in ANSI/ANS-18.1-1999, and the result should be corrected by reflecting the design characteristics of SMR. The coolant source term of NuScale, unlike the case of large NPPs, cannot rely solely on empirical source term data, because the NuScale source term is based on first principle physics, operational experience from recent industry, and lessons learned from large PWR operation. Fission products in reactor coolant are conservatively calculated using first principle physics in SCALE Code assuming 60 GWD/MTU. The release of fission products from fuel to primary coolant based on industry operational experience is determined as fuel failure fraction of 0.0066% for normal operation source term and 0.066% for design basis source term while coolant source term of large NPP is calculated by using ANSI/ANS-18.1 for normal operation and fuel failure fraction of 1% for design basis source term. Water activation products in reactor coolant are calculated from first principles physics and corrosion activation products are calculated by utilizing current large PWR operating data (ANSI/ANS 18.1- 1999) and adjusted to NuScale plant parameters. Also, because ANSI/ANS 18.1-1999 is not based on first principle physics models for CRUD generation, buildup, transport, plate-out, or solubility, NuScale has incorporated lessons learned by using ERPI’s primary water chemistry and steam generator guidelines to ensure source term is conservative and design of materials used cobalt reduction philosophy to help ensure the coolant source term are conservative. Based on the coolant source term calculated according to the above-described method, the annual releases of radioactive materials in gaseous and liquid effluents from NuScale reactor are evaluated. Currently, Small Modular Reactors such as ARA, SMART 100 are under review for licensing in Korea. This study will be helpful to understand how the reactor coolant system source terms are defined and evaluated for SMR.

In this study, a manual that can be applied to conflict management of clearance waste recycling by stakeholders was researched to recycle clearance waste that is most frequently generated when decommissioning nuclear power plants. In order to develop a manual that can be applied to conflict management, the content of the conflict should be derived first. In order to derive conflict, it is necessary to organize major issues in recycling clearance waste in consideration of domestic nuclear energy and social environment. In order to organize major issues in consideration of the domestic environment, a literature survey and a domestic current situation investigation were conducted. At this time, the subject of the major issue was selected based on the Level 1 influencing factors of the previous study. As a result of the investigation, it was confirmed that there were many major issues due to lack of reliability/understanding in nuclear energy/radiation. Through this Conflicts caused by recycling clearance waste were derived based on the organized issues. As a result of deriving conflicts, eight conflicts were derived below. 1) Reduced business availability due to lack of understanding/reliability 2) Lack of reliability in the selection and technology of nuclide analysis technology 3) Additional time and equipment required due to establishment of clearance waste regulatory requirements 4) Low economic benefits due to reduction in the effect of substituting raw materials 5) Political interference due to worsening public opinion 6) Rejection of final products due to recycling due to distrust of radiation 7) Public acceptance along the transport route from the source to the recycling plant 8) Business promotion deteriorated due to changes in energy policy As a result of the derived conflict analysis, the most conflicts related to lack of reliability/understanding in nuclear energy/radiation were derived. Accordingly, in future research, it is necessary to prepare a specific plan to enhance the understanding of stakeholders about self-disposal waste recycling. Considering that research that can solve the conflicts that will be faced when the domestic/foreign clearance waste recycling industry is activated is not activated, this study is meaningful in that it derived the conflicts that will be faced when recycling clearance waste. Also, it is expected that the conflicts derived from this study will be used meaningfully in the establishment of the clearance waste recycling management manual.

This study presents an example of creating and optimizing a task sequence required in an automated remote dismantling system using a digital manufacturing system. An automated remote dismantling system using a robotic arm has recently been widely studied to improve the efficiency and safety of the dismantling operations. The task sequence must be verified in advance through discrete eventbased process simulation in a digital manufacturing system to avoid problems in actual remote cutting operations as the main input of the automated remote dismantling system. The laser cutting method can precisely cut complicated target structures such as reactor internals with versatility, but a robot and a pre-prepared program are required to deploy sophisticated motion of the laser cutting head on the target structure. For safe and efficient dismantling operations, the robot’s program must be verified in advance in a virtual environment that can represent the actual dismantling site. This study presents creating and optimizing the task sequence of a robotic underwater laser cutting as part of the project of developing an automated remote dismantling system. A task sequence is created to implement the desired cutting path for the target structure using the automated remote dismantling system in the virtual environment. The task sequence is optimized for the posture of the laser cutting head and the robot to avoid collisions during the operation through discrete event-based process simulation since the target structure is complicated and the volume occupied by the laser cutting head and the robot arm is considerably large. The task sequence verified in the digital manufacturing system is demonstrated by experiments cutting the target structure along the desired cutting path without any problems. The various simulation cases presented in this study are expected to contribute not only to the development of the automated remote dismantling system, but also to the establishment of a safe and efficient dismantling process in the nuclear facility decommissioning.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

In the pilot scale test, the two scale-up factors (Electric energy per order EEO, Electric energy per mass EEM) were conducted to design the Chemical Waste Decomposition & Treatment System (CWDS). The CWDS consist of two kind UV lamp reactors to improve the decomposition rate of oxalic acid, which are low pressure amalgam UV lamp and medium pressure UV lamp. The two reactors were connected in series, and the hydrogen peroxide is mixed through a line mixer at the front of the reactor and injected into the reactors. The CWDS was connected with the full system decontamination equipment to purify the residual oxalic acid after chemical decontamination process. The full system decontamination equipment were included Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS) to operate the Oxidation/Reduction decontamination process and purify the process water. After decontamination process, the waste water will be cooled down into the 40°C and passed through the UV reactor at 110 gpm with hydrogen peroxide injection. The concentration of waste water is expected oxalic acid 1,700 ~ 2,000 ppm, Iron 5 ~ 20 ppm. As a result of the CBD test in the laboratory with simulated waste liquid, the amount of Low pressure amalgam lamp UV dose required to decompose 95% of oxalic acid in 2 m2 waste water was up to 1,800 mJ/cm2. The amount of medium pressure lamp UV dose was up to 450 mJ/cm2 at the same condition. We conducted demonstration test using 2 m2 waste water after the oxidation/reduction decontamination process, the decomposition rate 95% was obtained by low pressure amalgam UV lamp and medium pressure UV lamp reactor each.

The decommissioning of Kori Unit 1 is expected to generate a large amount of clearance waste. Disposing of a large amount of clearance waste is economically costly, so a recycling method has emerged. However, clearance waste recycling is expected to cause many conflicts among various stakeholders. In the previous study, possible conflicts were selected in consideration of the domestic environment and major issues. Based on this, this study classifies stakeholders involved in conflicts by group, and suggests ways to enhance understanding by stakeholder and enhance reliability. In this study, stakeholders are classified into four groups that share the same conflicts, and each of the following measures is suggested. 1) Stakeholder Engagement. 2) Common understanding of radiation risks, dialogue between the public/recycling industry/ regulatory agency. 3) Incentives to promote recycling clearance waste. 4) Reliable outlet store for recyclable clearance waste. The above understanding enhancement measures are presented so that a solution to conflict can be smoothly derived when designing a clearance waste-related consultative body composed of interested parties in the future. As a more specific solution, measures to enhance stakeholder trust can be suggested for each understanding enhancement measure. Reliability enhancement measures are also presented so that they can be applied to each stakeholder group, and these are as follows. 1) Write a stakeholder engagement plan, Measures for stakeholder participation in measuring the radioactivity concentration of clearance waste. 2) Active use of easy-to-understand radioactivity comparison data, Expansion of information on environmental radiation dose to public, nuclear/radiation education, Held a tour event at the nuclear power plant decommissioning site, New website for clearance waste information disclosure. 3) Incentives for recycling industries in which the Ministry of Environment or KHNP partially bears the losses that occur when the sales rate is low. Incentives are provided to consumers by including recyclables of clearance waste for Green Card’s green consumption points. 4) Online outlets open for recyclable clearance waste with easy-to-understand radioactivity comparison data. It is expected that if the above-mentioned reliability enhancement measures are used, it will be possible to secure the trust of stakeholders and reduce the gap between stakeholders in the future clearance-related consultative body.