Radiation workers, especially those dealing with Uranium isotopes, can potentially intake Uranium -containing materials through their respiratory and digestive systems. According to the “Regulations on the Measurement and Calculation of Internal Exposure” from Nuclear Safety and Security Commission (NSSC), those who intend to work in or enter the nuclear facilities with a risk of exceeding 2 mSv exposure per year should be examined the internal exposure. However, when it comes to in-vitro bioassay, Uranium intake through drinking water can affect the quantitative analysis. The International Commission on Radiological Protection (ICRP) reported in ICRP Publication 23 (Report on the Task Group on Reference Man) that the reference man excretes Uranium in the urine (0.05-0.5 μg/day) and feces (1.4-1.8 μg/day). Korea Atomic Energy Research Institute (KAERI) set the 90.5 ng/day as the 238U background of workers handing Uranium based on the daily Uranium intake of Koreans. In this research, we examined the possible effects of Uranium in drinking water on internal exposure by analyzing the concentration of Uranium in bottled waters from various water sources sold in the domestic market and a water from the water purifier. The 238U concentration results of analyzing 11 bottled waters and 1 purified water, were ranged from 0 to 10.2 μg/L. All the results were satisfied the standard of 30 μg/L according to “Regulations for Drinking Water Quality Standards and Inspection” enacted by the Ministry of Environment. However, various concentrations were shown depending on the water sources. Assuming that these concentrations of water are consumed by drinking 1 L per day, the internal dose assessment result is 0 to 0.94 mSv. On the other hand, if it is assumed to be inhaled, it can be an overestimated because the dose coefficient of inhalation, Type M is higher than that of ingestion, f1=0.02 which are the values recommended by ICRP Publication 78 (Individual Monitoring for Internal Exposure of Workers) when the Uranium compound is unspecified. In case of two workers at KAERI, the daily excretion of urine was 151 and 120 ng/day respectively in the first quarter monitoring. However after changing the kind of drinking water in the second quarter monitoring, it dropped to 17.4 and 15.4 ng/day respectively. Through this study, it is confirmed that the Uranium background in urine can be analyzed differently depending on the kind of drinking water consumed by each worker. Depending on the Uranium concentration of drinking water, the internal exposure dose assessment can be overestimated or underestimated. Therefore, the Uranium concentration and intake amount according to the kind of drinking water should be considered for in-vitro bioassays of Uranium handlers. Furthermore, if necessary, the Uranium isotope ratio analysis in urine and the handling information should be comprehensively considered. In addition, in order to exclude the effect of intake through the digestive system, replacing the kind of drinking water can be considered. The additional analysis such as in-vivo bioassay and 24 hours urine analysis rather than spot samples can be also recommended.

To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

If radioactive plumes are released outside due to loss of containment building integrity during a nuclear power plant accident, these materials might travel with the wind, affecting both the surrounding environment and neighboring countries. In China, most nuclear power plants are located on the eastern coast. Consequently, a radioactive plume generated during an accident could negatively impact even the western part of the Korean Peninsula due to westerly winds. To detect such problems early, respond quickly, and protect residents, a system that can monitor aerial radiation under normal conditions is needed. Additionally, a detection system that can operate in real-time in an emergencies conditions is required. The current method for aerial radiation measurement takes environmental radiation data from a monitoring post 1.5 m above the ground and converts it to altitude. To measure actual aerial radiation, an expansive area is surveyed by aircraft. However, this approach is both time-consuming and expensive. Thus, to monitor radioactive plumes influenced by environmental factors like wind, we need a radiation detector that can gauge both radioactivity and directionality. In this study, we developed a radiation detector capable of assessing both the radioactivity and directionality of a radioactive plume and conducted its performance evaluation. We miniaturized the radiation detector using a CZT (Cadmium Zinc Telluride) sensor, enabling its mounting on unmanned aerial vehicles like drones. It is configured with multi-channels to measure directionality of a radioactive plumes. For performance evaluation, we positioned two-channel CZT sensors at 90 degrees and measured the energy spectrum for angle and distance using a disk-type radioactive isotope. Using this method, we compared and analyzed the directionality performance of the multi-channel radiation detector. We also confirmed its capability to discern specific radioactivity information and nuclide types in actual radioactive plumes. Our future research direction involves mounting the multi-channel radiation detector on a drone. We aim to gather actual aerial radiation data from sensors positioned in various directions.

With an ultimate view to identifying abnormal releases of radioactive materials, a set of liquid and gaseous effluent data including unplanned or uncontrolled releases annually reported form the U.S. and Korean nuclear power plants were systematically analyzed. With the use of 21 years’ worth of annual discharge data for 7 radionuclide groups and 24 individual radionuclides, taken from a combined total of 1,610 reactor-years (RYs) covering 62 units of US Pressurized Water Reactors (PWRs) and 22 units of Korean PWRs, three novel formulas for estimating events were employed to calculate characteristic values. Applying these characteristic values derived from the event estimation formulas to events that transpired during 699 RYs in operational US PWRs revealed an enhanced predictive accuracy for abnormal events when considering individual radionuclides, as opposed to grouping them by radionuclide groups. This effect was particularly pronounced for specific events such as leaks caused by problems in Gas Decay Tanks, leaks in Steam Generator Power Operated Relief Valves, fuel defects, and leaks during spent nuclear fuel processing. In the case of Korean PWRs, fuel defects were identified as the primary events related to radioactive effluent releases. The methodologies and characteristic values derived from this study were applied to these events. The event estimation rate was lower in Korean compared to US PWRs, which can be attributed to the lower frequency of event occurrences in Korean PWRs (30 RYs) compared to the US. The approach proposed in this study may contribute to develop a methodology to identify implicit abnormal release data and correlate them with specific operational occurrences or events, which could improve the conventional practice of simply recording and reporting radioactive discharge data.

This study presents distribution of naturally occurring radioactive materials in groundwater in Jeju island. Radon (222Rn) and potassium (40K) concentrations were performed by using Liquid Scintillation Counter and Ion Chromatograph respectively. In addition, the activities of uranium and thorium nuclides were analyzed by Inductively Coupled Plasma Mass Spectroscopy. Groundwater samples were collected from 9 sites of water intake facilities for wide area supply in Jeju island from September 2022 to September 2023. The 40K concentrations of groundwater ranged between 0.050 and 0.400 Bq·L-1. The radon concentrations in groundwater were in the range of 0 to 60 Bq L-1, and there was no groundwater exceeding the range of 148 Bq L-1 proposed by the US EPA. The distribution of uranium and thorium in groundwater varied from 0 to 500 ng L-1 and 0 to 2.4 ng L-1, respectively. The concentrations of uranium did not exceed 30 μg L-1, thresholds indicated by the US EPA. By analyzing the concentrations of 40K, 222Rn, 238U and 232Th, the annual effective dose of residents can be assessed. The evaluated residents’ effective dose from natural radionuclides due to intake of drinking water is less than the recommended value of 100 μSv y-1. Consequently, this study indicates that the cancer risks of the residents in Jeju island from naturally occurring radioactive materials ingested with water is insignificant.

Copper hexacyanoferrate (Cu-HCF), which is a type of Prussian Blue analogue (PBA), possesses a specific lattice structure that allows it to selectively and effectively adsorb cesium with a high capacity. However, its powdery form presents difficulties in terms of recovery when introduced into aqueous environments, and its dispersion in water has the potential to impede sunlight penetration, possibly affecting aquatic ecosystems. To address this, sponge-type aluminum oxide, referred to as alumina foam (AF), was employed as a supporting material. The synthesis was achieved through a dip-coating method, involving the coating of aluminum oxide foam with copper oxide, followed by a reaction with potassium hexacyanoferrate (KHCF), resulting in the in-situ formation of Cu-HCF. Notably, Copper oxide remained chemically stable, which led to the application of 1, 3, 5-benzenetricarboxylic acid (H3BTC) to facilitate its conversion into Cu-HCF. This was necessary to ensure the proper transformation of copper oxide into Cu-HCF on the AF in the presence of KHCF. The synthesis of Cu-HCF from copper oxide using H3BTC was verified through X-ray diffraction (XRD) analysis. The manufactured adsorbent material, referred to as AF@CuHCF, was characterized using Fourier-transform infrared spectroscopy (FTIR) and thermogravimetric analysis (TGA). These analyses revealed the presence of the characteristic C≡N bond at 2,100 cm-1, confirming the existence of Cu-HCF within the AF@CuHCF, accounting for approximately 3.24% of its composition. AF@CuHCF exhibited a maximum adsorption capacity of 34.74 mg/g and demonstrated selective cesium adsorption even in the presence of competing ions such as Na+, K+, Mg2+, and Ca2+. Consequently, AF@CuHCF effectively validated its capabilities to selectively and efficiently adsorb cesium from Cs-contaminating wastewater.

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.

The thermal treatment of radioactive waste attracts great attention. The thermal treatment offers lots of advantages, such as significant volume reduction, hazard reduction, increase of disposal safety, etc. There are various thermal technologies to waste. The developed technologies are calcination, incineration, melting, molten salt oxidation, plasma, pyrolysis, synroc, vitrification, etc. The off-gas treatment system is widely applied in the technologies to increase the safety and operation efficiency. The thermal treatment generates various by-product and pollutants during the process. The dust or fly ash are generated as a particulate from almost every radioactive waste. The treatment of PVC related components generates hydrogen chloride, which usually brings corrosion of facility. The treatment of rubber and spent resin generates sulfur oxide, SOx. The treatment of nitrile rubber generates nitrogen oxide, NOx. The incomplete combustion of radioactive waste usually generates carbon oxide, COx. The process temperature also affects the generation of off gas, such as NOx and/or COx. Various off gas treatment components are organized for the proper treatment of the previously mentioned materials. In this study systematical review on off gas treatment will be reported. Also, worldwide experiences and developed facility will be reported.

The primary purpose of high temperature process of radioactive waste is to satisfy the waste acceptance criteria and volume reduction. The WAC offers the guideline of waste form fabrication process. The WAC is defined as quantitative or qualitative criteria specified by the regulatory body, or specified by and operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. The main objective of WAC is to protect staff and general public and environment by the containment of radioactive material, limit external radiation level, and prevent criticality. The WAC also offers systematic management of radioactive waste by standardization of waste management operations, facilitation waste tracking, ensure safe and effective operation of operating facilities, etc. Since the high temperature process for radioactive waste is considered in many countries, lots of codes and standards are considered. In many WACs, compressive strength, thermal cycle stability, radiation exposure stability, free liquid, and leachability are evaluation to understand the effect of solidified form to the disposal facility. In this paper, systematical review on waste form will be discussed. In addition, brief result of characterization of waste form will be compared.

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

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.

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.

For safe and successful decommissioning, it is one of the most important procedures that establishing the goal and complying with regulations of which final status of decommissioned site and building. The dose criteria for cyclotron facilities should be established and applied to reuse the site and building, since building and component of a cyclotron facility have been activated by incident secondary neutrons from radioactive isotope processes (e.g. 18O(p,n)18F, etc.). Furthermore, appropriate approaches should be applied to demonstrate compliance with the dose criteria for reliability of reuse. It is of noted that U.S. NRC (Nuclear Regulatory Commission) has confirmed that the residual radioactivity which distinguishable from background radiation results in a TEDE (Total Effective Dose Equivalent) does not exceed 25 mrem (0.25 mSv) per year as radiological criteria for unrestricted use of not only nuclear power plants but also cyclotron facilities referred to 10 CFR Part 20.1402. In addition, U.S. NRC noted the two approaches (i.e. dose assessment methods and, DCGL and final status surveys) which can be applied for demonstrating compliance with the dose criteria of 10 CFR Part 20 and recommended DCGL and FSS approach based on advantages and disadvantages of the two approaches. In order to using DCGL and FSS approach, U.S. NRC suggested screening approach; using DandD Version 2 which assesses TEDE under ICRP 28 and site-specific approach; using all models or computational codes which approved by NRC staff. There are several foreign cases that release of cyclotron facilities after decommissioning (i.e. U.S. and Japan). U.S., for examples, there are two DCGL approach cases and one dose modeling case based on 25 mrem per year same as reactor facilities. The dose modeling case, however, which may not be really used in Korea because of its low applicability. On the other hand, Japan case did not establish any radiological criteria for site and building reuse such as DCGL and just confirm “no more contamination” which is all residual radioactivity is lower than MDC based on real survey. Japan case also may not be used in Korea since criteria of “no more contamination” is not clear and hard to apply for all sites. Considering regulations and criteria for site release and reuse in Korea, this study aims to suggest radiological criteria and the demonstration approach of compliance for decommissioning of cyclotron facilities based on Nuclear Safety Acts and NSSC notices.

Pt/C catalysts were prepared using black carbon (CB), and evaluated for their potential application as a catalyst of liquid-phase catalystic exchange for tritium treatment. CB was treated with 10% H2O2 solution for 0 and 2 hours at 105°C, Ethylene glycol and 40wt% Pt were added to the dried treated sample to prepare a Pt/C catalyst. The physical and chemical properties of the prepared catalysts were evaluated by BET, XRD, elemental analysis (EA), and TEM analyses. As a result of BET analysis, the surface area of CB without 10% H2O2 was 237.2 m2·g-1, and after treatment with 10% H2O2, it decreased to 181.2 m2·g-1 for 2 hours. However, the internal surface area increased, indicating the possibility that more Pt could be distributed inside the CB treated with 10% H2O2. In the XRD analysis results, the presence of Pt was confirmed by observing the Pt peak in the prepared Pt/C catalyst, and it was also observed through TEM analysis that Pt was evenly distributed within the CB. The elemental analysis (EA) results showed that the ratio of S and N decreased and the ratio of O increased with increasing 10% H2O2 treatment time. The H2O2 treated carbon supported Pt catalysts and polytetrafluoroethylene were then loaded together on a foamed nickel carrier to obtain hydrophobic catalysts. Our hydrophobic Pt catalyst using H2O2 treated black carbon are expected to be usefully used in the tritium treatment system.

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.

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.

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.

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.

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).

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.