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.

Nuclear facilities present the important task related to the migration and retention of radioactive contaminants such as cesium (Cs), strontium (Sr), and cobalt (Co) for unexpected events in various environmental conditions. The distribution coefficient (Kd) is important factor for understanding these contaminants mobility, influenced by environmental variables. This study focusses the prediction of Kd values for radionuclides within solid phase groups through the application of machine-learning models trained on experimental data and open source data from Japan atomic energy agency. Three machine-learning models, such as the convolutional neural network, artificial neural network, and random forest, were trained for prediction model of the distribution coefficient (Kd). Fourteen input variables drawn from the database and experimental data, including parameters such as initial concentration, solid-phase characteristics, and solution conditions, served as the basis for model training. To enhance model performance, these variables underwent preprocessing steps involving normalization and log transformation. The performances of the models were evaluated using the coefficient of determination. These results showed that the environmental media, initial radionuclide concentration, solid phase properties, and solution conditions were significant variables for Kd prediction. These models accurately predict Kd values for different environmental conditions and can assess the environmental risk by analyzing the behavior of radionuclides in solid phase groups. The results of this study can improve safety analyses and longterm risk assessments related to waste disposal and prevent potential hazards and sources of contamination in the surrounding environment.

The process of carbonization followed by a high-temperature halogenation removal of radionuclides is a promising approach to convert low-radioactivity spent ion-exchange (IE) resins into freereleasable non-radioactive waste. The first step of this process is to convert spent ion-exchange resins into the carbon granules that are stable under high-temperature and corrosive-gas flowing conditions. This study investigated the kinetics of carbonization of cation exchange resin (CER) and the changes in structures during the course of carbonization to 1,273 K. Both of model-free and modelfitted kinetic analysis of mixed reactions occurring during the course of carbonization were first conducted based on the non-isothermal TGAs and TGA-FTIR analysis of CER to 1,272 K. The structural changes during the course of carbonization were investigated using the high-resolution FTIR and C-13 NMR of CER samples pyrolyzed to the peak temperature of each reaction steps established by the kinetic analysis. Four individual reaction steps were identified during the course of carbonization to 1,273 K. The first and the third steps were identified as the dehydration and the dissociation of the functional group of —SO3-H+ into SO2 and H2O, respectively. The second and the fourth steps were identified as the cleavage of styrene divinyl benzene copolymer and carbonization of pyrolysis product after the cleavage, respectively. The temperature and time positions of the peaks in the DTG plot are nearly identical to those of the peaks of the Gram Schmidt intensity of FTIR. The structural changes in carbonization identified by high-resolution FTIR and DTG are in agreement with those by C-13 NMR. The results of a detailed examination of the structural changes according to NMR and FTIR were in agreement with the pyrolysis gas evolution characteristics as examined by TGA-FTIR.

Hydrogen isotope separation involves the separation of hydrogen, deuterium, tritium, and their isotopologues. It is an essential technology for removing radioactive tritium contamination and for obtaining valuable hydrogen isotope resources. Among various hydrogen isotope separation technologies, water electrolysis technology exhibits a high separation factor. Consequently, the electrolysis of tritiated water is of paramount importance as a tritium enrichment method for treating tritium-contaminated water and for analyzing tritium in environmental samples. More recently, hydroelectrolysis technology, which utilizes proton exchange membranes (PEM) to reduce water inventory, has gained favor over traditional alkaline hydroelectrolysis. Nevertheless, it is crucial to decrease the hydrogen permeability of the PEM in order to mitigate the explosion risk associated with tritium hydrogen electrolysis devices. Additionally, efforts are needed to enhance the hydrogen isotope selectivity of the PEM and optimize the manufacturing process of the membrane-electrode assembly (MEA), thereby improving both hydrogen isotope separation performance and water electrolysis efficiency. In this presentation, we will delve into two key aspects. Firstly, we’ll explore the reduction of hydrogen permeability and the enhancement of the hydrogen isotope separation factor in PEM through the incorporation of 2D nanomaterial additives. Secondly, we’ll examine the influence of various MEAs preparation methods on electrolysis and isotope separation performances. Lastly, we will discuss the effectiveness of the developed system in separating deuterium and tritium.

Despite its advantages such as safety, unnecessary pretreatment, and decontamination of waste with complex geometry, conventional ultrasonic decontamination technology has been only used to remove loose contaminants, oil and grease, not fixed contaminants due to the limitations in increasing the intensity in the high frequency range. Thus, ultrasound has been used as an auxiliary method to accelerate chemical decontamination of radioactive wastes or chemicals were added to the solution to increase the decontamination efficiency. The recently developed high-intensity focused ultrasound (HIFU) decontamination technology overcomes these limitations by combining multiple frequencies of ultrasonic waves in a specific arrangement, making it possible to remove most fixed contaminants, including radioactive micro particles less than 1 micrometer within half an hour. KEPCO NF and EnesG developed mobile HIFU decontamination equipment and successfully demonstrated the decontamination effect on various radionuclides found in nuclear power plants by treating radioactive metal waste to the level below free release criteria. The mobile HIFU decontamination equipment used in the demonstration can be operated anywhere where water is supplied, including controlled area in nuclear power plants, and is expected to be used widely for decontamination and free release of metal radioactive wastes.

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).

Economical radioactive soil treatment technology is essential to safely and efficiently treat of high-concentration radioactive areas and contaminated sites during operation of nuclear power plants at home and abroad. This study is to determine the performance of BERAD (Beautiful Environmental construction’s RAdioactive soil Decontamination system) before applying magnetic nanoparticles and adsorbents developed by the KAERI (Korea Atomic Energy Research Institute) which will be used in the national funded project to a large-capacity radioactive soil decontamination system. BERAD uses Soil Washing Process by US EPA (402-R-007-004 (2007)) and can decontaminate 0.5 tons of radioactive soil per hour through water washing and/or chemical washing with particle size separation. When contaminated soil is input to BERAD, the soil is selected and washed, and after going through a rinse stage and particle size separation stage, it discharges decontaminated soil separated by sludge of less than 0.075 mm. In this experiment, the concentrations of four general isotopes (A, B, C, and D which are important radioisotopes when soil is contaminated by them.) were analyzed by using ICP-MS to compare before and after decontamination by BERAD. Since BERAD is the commercial-scale pilot system that decontaminates relatively large amount of soil, so it is difficult to test using radioactive isotopes. So important general elements such as A, B, C, and D in soil were analyzed. In the study, BERAD decontaminated soil by using water washing. And the particle size of soil was divided into a total of six particle size sections with five sieves: 4 mm, 2 mm, 0.850 mm, 0.212 mm, and 0.075 mm. Concentrations of A, B, C, and D in the soil particles larger than 4 mm are almost the lowest regardless of before and after decontamination by BERAD. For soil particles less than 4 mm, the concentrations of C and D decreased constantly after BERAD decontamination. On the other hand, the decontamination efficiency of A and B decreased as the soil particle became smaller, but the concentrations of A and B increased for the soil particle below 0.075 mm. As a result, decontamination efficiency of one cycle using BERAD for all nuclides in soil particles between 4 mm and 0.075 mm is about 45% to 65 %.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

When aluminum is in an alkaline state, the aluminum oxide film surrounding aluminum is dissolved and moisture penetrates the exposed aluminum surface, causing corrosion of aluminum. At this time, hydrogen gas is generated and there is a risk of explosion due to the generated hydrogen gas. Aluminum radioactive waste is difficult to permanently dispose of because it does not meet the standards for the acquisition of low- and intermediate-level radioactive waste cave disposal facilities currently managed and operated by the Korea Nuclear Environment Corporation. However, because of this risk, it is necessary to study how to safely treat and dispose aluminum waste. In this study, overseas cases were investigated and analyzed to ensure the safety of aluminum waste disposal, and the current status of aluminum radioactive waste generated during decommissioning of the Korea Research Reactor 1&2 and a treatment plan to secure disposal suitability were presented. The process of removing a little remaining oxygen in molten steel during the reduction of iron oxide in the iron refining process is called deoxidation, and a representative material used for deoxidation is aluminum. In the case of metal melting decontamination, which is one of the decontamination processes of radioactive metal waste, a method of treating aluminum waste by using aluminum as a deoxidizer is proposed.

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

The periodic safety review (PSR), for all operating nuclear power plants in Korea, has been conducted in accordance with SSG-25, a guideline suggested by the IAEA, The PSR is performed through the review of the regulatory body after the operator’s self-evaluation. In order to guarantee a high level of safety in consideration of the changed environment, such as operating experience (OE) and technology development, it should be comprehensively and integratedly performed, and it is also carried out every 10 years after the operation permit. However, in case that all or part of the reactor facilities have been permanently shut down, such as Kori Unit 1 and Wolsong Unit 1, Around a half of reactor facilities are not in operation. The periodic safety evaluation may not be conducted for unused parts if there is no safety hazard and if there are some difficulties for applying periodic safety evaluation. In considering that the biggest purpose of PSR safety (by PSR definition of KINS guideline) is to improve and accumulated factors such as aging deterioration, facility change, operation experience, and technological development for operating nuclear power plants. It refers to a comprehensive safety evaluation that is periodically performed during the period of operation of a nuclear power plant. It is necessary to review whether PSR should be performed for a nuclear power plant that is permanently shut down after nuclear power plant operation is terminated. Also, in IAEA SSR 2/2 Rev1, it is defined that PSR is performed during the nuclear power plant operation period. “Requirement 12: Periodic safety review, Systematic safety assessments of the plant, in accordance with the regulatory requirements, shall be performed by the operating organization throughout the plant’s operating lifetime, with due account taken of operating experience and significant new safety related information from all relevant sources”. Recently, Kori Unit 1 and Wolsong Unit 1 were decided to permanently shut down in June 2017 and December 2019, and are currently being prepared for decommissioning. According to the Wolsong decommissioning plan, decontamination and demolition will be completed by 2032. The PSR for permanent shutdown of Kori Unit 1 was submitted to the regulatory body in December 2018 and is under approval review. In the case of the permanent shutdown PSR of Wolsong Unit 1, the project will be launched in May 2023 and the PSR will be submitted to the regulatory body in May 2024. In the case of Wolsong Unit 1, it is necessary to operate the various systems, including the systems related to the spent fuel storage tank, even during the period of permanent shutdown. Such as the heavy water related systems used in common with Wolsong Unit 2, are essential operating systems. Based on Basic Subject Index (BSI), 112 out of 218 systems require operation, indicating that about 50% of systems require operation even after permanent shutdown. Decommissioning of systems and equipment will begin after the transfer to modular air-cooled canister storage (MACSTOR) by the end of 2025, and then in-depth discussions will be needed whether PSR evaluation is meaningful.

Decommissioning waste is generated with various types and large quantities within a short period. Concrete, a significant building material for nuclear facilities, is one of the largest decommissioning wastes, which is mixed with aggregate, sand, and cement with water by the relevant mixing ratio. Recently, the proposed treatment method for volume reduction of radioactive concrete waste was proven up to scale-up testing using unit equipment, which involved sequentially thermomechanical and chemical treatment. According to studies, the aggregate as non-radioactive material is separated from cement components with contaminated radionuclides as less than clearance criteria, so the volume of radioactive concrete waste is decreased effectively. However, some supplementation points were presented to commercialize the process. Hence, the process requires efficiency as possible to minimize the interface parts, either by integration or rearranging the equipment. In this study, feasibility testing was performed using integrated heating and grinding equipment, to supplement the possible issue of generated powder and dust during the process. Previously, heat treatment and grinding devices were configured separately for pilot-scale testing. But some problems such as leakage and pipe blockage occurred during the transportation of generated fine powder, which caused difficulties in maintaining the equipment. For that reason, we studied to reduce the interface between the equipment by integrating and rearranging the equipment. To evaluate the thermal grinding performance, the fraction of coarse and concrete fines based on 1mm particle size was measured, and the amount of residual cement in each part was analyzed by wet analysis using 4M hydrochloric acid. The result was compared with previous studies and the thermomechanical equipment could be selected to enhance the process. Therefore, it is expected that the equipment for commercialization could be optimized and composed the process compactly by this study.

Working with molten metal has always been and will always be a dangerous workplace. No matter how carefully equipment is designed, workers are trained and procedures are followed, the possibility of an accident can occur in melting workplace. Some primary causes of melt splash and furnace eruptions include wet or damp charge material, dropping heavy charge into a molten bath, wet or damp tools or additives and sealed scrap or centrifugally cast scrap rolls. Induction melting brings together three things (water, molted metal and electricity) that have the potential for concern if the furnace is not properly working. Induction furnace must have a water cooling system built into the coil itself. Water picks up the heat caused by the current as well as heat conducted from the metal through the refractory. The water carries the heat to a heat exchange for removal. Spill pits serve to contain any molten metal spilled as a result of accident, run out or dumping of the furnace in an emergency. If a leak is suspected at any time, cease operation and clear the melt deck area of all personnel and empty the furnace. Molten metal fins can penetrate worn or damaged refractory and come into contact with the coil. A furnace or a close capture hood which suddenly swings down from a tilted position will cause injury or death. Whenever workers are working on a furnace or close capture hood when it is in the tilted position, be sure that it is supported with a structural brace that is strong enough to keep it from dropping if hydraulic pressure is lost. In theory refractory wear should be uniform, however, in practice this never occurs. The most causes of lining failure are improper installation of refractory material, inadequate sintering of refractory material, failure to monitor and record normal lining wear, allowing the lining to become too thin, installation of the wrong refractory, improper preheating of a used cold lining, failure to properly maintain the furnace the sudden or cumulative effects of physical shocks or mechanical stress, and excessive slag or dross buildup. Pouring cradles provide bottom support for crucibles. A crack in the crucible occur below the bottom ring support, the bottom of the crucible can drop and molten metal will spill and splash, possibly causing serious injury or death. To reduce this danger, a pouring cradle that provides bottom support for the crucible must be used. Power supply units must have safety locks and interlocks on all doors and access panels. Workers who work with low voltage devices must be made aware of the risk posed by high levels of voltage and current. The most causes of accidents are introduction of wet or damp material, improper attention to charging, failure to stand behind safety lines, coming into contact with electrically charged components and lack of operator skills and training. Only trained and qualified personnel are to have access to high risk areas. Safety lockout systems are another effective measure to prevent electrical shock

The critical hazards generated from operation of a melting facility for metal radioactive waste are mainly assumed to be such as vapor explosion, ladle breakthrough and failure in the hot-cell or furnace chamber using remote equipment. In case of vapor explosion, material containing moisture and/or enclosed spaces may, due to rapid expansion of gases when heated, cause an explosion and/or violent boiling. The rapid expansion of gases may lead to ejection of molten radioactive metal from the furnace into the furnace hall. If there is a large amount of liquid the explosion may damage or destroy technical barriers such as facility walls. The consequences for the facility ranges from relatively mild to very severe depending on the force of the explosion as well as the type of waste being melted. Nonradiological consequences may be physical damage or destruction of equipment and facility barriers, such as walls. Due to the radiological consequences a longer operational shutdown would likely be required. Cleanup efforts would include cutting of solidified metal in a problematic radiological environment requiring use of remote technology before damage and repair requirements can be assessed. Even though there is a risk for direct physical harm to operators for example in the control room and hot-cell, this analysis focuses mainly on the radiological impact. The extent to which remote equipment could be used in the decontamination effort will largely determine the health consequences to the workers. It is reasonable to assume that there will be a need for workers to participate manually in the effort. Due to the potentially large dose rates and the physical environment, it is possible that the maximum allowable dose burden to a worker will be reached. No major consequence for the environment is expected as most of the radioactivity is bound to the material. In case of ladle breakthrough, a ladle breakthrough involves loss of containment of the melt due to damage of the ladle. This may be caused e.g. by increased wear due to overheating in the melt, or from physical factors such as mechanical stress and impact from the waste. A ladle breakthrough may lead to spread of molten metal in the furnace hall. Molten metal coming into contact with the surrounding cooling equipment may cause a steam explosion. The consequences of a ladle breakthrough will depend on the event sequence. The most severe is when the molten metal comes into contact with the cooling system causing a vapor explosion. The basic consequences are assumed to be similar to those of the vapor explosion above. While the ejection of molten metal is likely more local in the ladle breakthrough scenario, the consequences are judged to be similar. In case of failure in the hot-cell or furnace chamber using remote equipment, the loss of electric supply or technical failure in the furnace causes loss of power supply. If not remedied quickly, this could lead to that the melt solidifies. A melt that is solidified due to cooling after loss of power cannot be removed nor re-melted. This may occur especially fast if there is not melted material in the furnace. An unscheduled replacement of the refractory in the furnace would be required. It could be unknown to what degree remote equipment can be used to cut a solidified melt. It is therefore assumed that personnel may need to be employed. This event could not have any impact on environment

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

As unit 1 of Kori was permanently shut down in June 2017, domestic nuclear industry has entered the path of decommissioning. The most important thing in decommissioning is cost reduction. And volume reduction of radioactive waste is especially important. According to the IAEA report, more than 4,000 tons of metallic waste is generated during the decommissioning of a 1,000 MWe reactor and most of these wastes are LLW or VLLW. To reduce amount of metallic waste dramatically, we should choose efficient decontamination method. In this study, we conducted dry ice and bead blasting decontamination. We prepared Inconel-600 and STS-304 specimen with dimensions of 30 mm × 30 mm × 5 mm. Loose and fixed contamination was applied on the surface of specimen using SIMCON method. Bead and dry-ice blasting was conducted by spraying alumina and dry ice pellet at the same pressure and distance for the same time. The removal of loose contamination was observed using microscope. It was found that contaminants are significantly removed using both dry ice blasting and bead blasting. However, some abrasive material remained on the surface of specimen. The removal of fixed contamination was verified by weight comparison before and after experiment and cobalt concentration comparison before and after experiment using X-ray Fluorescence Spectroscope (XRF). At least 90% of the cobalt was removed, but some abrasive particle was also remained on the surface of specimen. In this study, it is confirmed that the effectiveness of manufacturing a large-scale abrasive decontamination facility, and it is expected that this technology can be used to effectively reduce the amount of metallic waste generated during decommissioning.

Chelate resin is a resin that has an exchange group which can form chelates with various metal ions. It shows higher selectivity for metal ions than ion exchange resin and can selectively remove characteristic metal ions. In an aqueous solution containing metal ions, chelate resin can adsorb specific metal ions, and the separated chelate resin can desorb the adsorbed metal ions by changing temperature or pH, so chelate resin has the advantage of being reusable. Chelate resin has been used industrially as an adsorbent to adsorb and separate heavy metal ions in wastewater, and is also used for the purpose of recovering precious or rare metals contained in industrial wastewater or industrial waste. Against this background, there is a need to develop chelate resins with higher adsorption capacity. Acrylic fiber is defined as a man-made fiber made from a linear synthetic polymer with fiberforming ability consisting of more than 85% acrylonitrile. It is a man-made fiber that is often used as a substitute for wool because it has good thermal insulation properties like wool and is warm and soft to the touch. It is a fiber rich in cyano groups due to its high content of acrylonitrile, and has the advantage of being able to be used as a variety of functional fibers through modification of cyano groups. In this study, the amination reaction of acrylic fiber was performed using diethylenetriamine, and the adsorption characteristics for metal ions were evaluated according to the reaction conversion rate. In order to improve the amination efficiency, 400 kGy was irradiated using a 2.5 MeV electron beam accelerator, and through this, the crosslinking rate of acrylic fiber was able to be improved up to 80%. Water and ethanol were used as cosolvents for the amination reaction in a ratio of 60/40 vol/vol, respectively, and a reaction yield of 178% was obtained after 120 minutes of reaction. Using the chelate resin prepared in this way, the adsorption performance for metal ions was evaluated through Atomic Absorption Spectrometry analysis.

A Partially hydrolyzed poly (vinyl acetate) (PHPVA)-borax complex-based gel-like coating was successfully developed for the decontamination of Simulated nuclear fallout (SFO) from surfaces. The sprayable coating was self-generated on the surface by borate-diol ester bonds after simultaneously mixing two solutions of borax and PHPVA. The SFO particles, synthesized at 1,200°C for melting, were glassy while some crystalline phases (e.g., SiO2 and Fe2O3) existed together. The SFO particles were fixed onto the Stainless steel (SS) substrate by dropping and evaporating water. for examination of the dust-removal performance of PHPVA-borax based coating. The dusts on the SS surface was successfully removed by casting the PHPVA-borax based coating within 1 minute, demonstrating the excellent dust-removal property of the PHPVA-borax based coating. The used PAB complex in wet state was recovered by using vacuum suction machine in short time. The solid-state PHPVA-borax based film was self-delaminated from the SS substrate after fully drying the used PHPVA-borax coating but this requires long period of time.

Wide-area surface decontamination is essential during the sudden release of radioisotopes to the public, such as nuclear accidents or terrorist attacks. A spray coating composed of a reversible complex between poly (vinyl alcohol) (PVA) and phenylboronic acid-grafted poly (methyl vinyl ether-alt-mono-sodium maleate) (PBA–g–PMVE–SM) was developed to remove radioactive cesium from surfaces. The simultaneous spay of PVA and PBA–g–PMVE–SM aqueous polymer solutions containing Cs adsorbent to contaminated surfaces resulted in the spontaneous formation of a PBA–diol ester bond-based gel-like coating. The Cs adsorbent suspended in the gel-like coating selectively removed Cs-137 from the Cs-contaminated surface. The used gel-like coating were removed from surfaces by simple water rinsing. This recovery way has advantages compared with costly incineration to remove the organic materials for final disposal/storage of the radioactive waste. Thus, our spray coating is suitable for practical wide-area surface decontamination. In radioactive tests, the hydrogel containing Cs-adsorbent showed substantial Cs-137 removal efficiencies of 96.996% for painted cement and 63.404% for cement, which are 2.33 times better than the values for the commercial surface decontamination coating agent DeconGel.

General phases in the plan and implementation of an environmental remediation of radioactively contaminated sites are planning for remediation, site characterization, remediation criteria, remediation strategy, implementing remediation actions, and conducting post-remediation activities. Environmental remediation should commence with a planning stage. It is helpful to prepare reports which detail all the supporting activities related to these elements before significant levels of funds and efforts are committed. Site characterization is needed to provide sufficient data to make strategic decisions on the environmental remediation activities. The source characterization should include both waste characterization and facility or site characterization and should provide reliable estimates of the release rates of radioactive constituents as well as constituent distribution. During the preliminary site characterization, an engineering study should be conducted to develop remediation options which address the specific contaminant problem and are aimed to reduce radiological and chemical exposure. Options will include engineering approaches and associated technologies. A preliminary selection of options may be made based on several factors including future sites use, technical considerations, public acceptability, cost, and regulatory requirements. The implementation of remediation actions includes procurement of the selected technology, preparation of the site, development of a health and safety plan, development of operations procedures, staff selection and training, completion of site cleanup, verification, waste disposal, and release of the site for any future use. Once remediation activities have been completed and verified, the remediated site can be released for restricted or unrestricted use. Remediation of radioactively contaminated sites may require special adaptation to address sites covering very large surface areas or those which are deep and difficult to access. Quality assurance may be very important to the verification of environmental remediation activities. The selection of optimal remediation technologies to solve or mitigate the safety of an environmental contamination problem should be taken into account several factors. The several factors include performance (the ability of the technology to reduce risk to the health and safety of the public and to the environment), reliability and maintenance requirements for the technology, costs of implementing the technology, infrastructure available to support the technology, availability(the ease of accessing the technology and associated services), risk to workers and public safety, environmental impacts of the technology, ability of the technology to meet regulatory acceptance, and communication of stakeholder.