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

LiCl-KCl eutectic possesses unique properties such as a low melting point, high thermal conductivity, and good electrical conductivity. These properties make it suitable for various applications, including nuclear power generation, pyroprocessing in nuclear waste management, and thermal energy storage systems. In most experiments using LiCl-KCl, the molten salt composition is an important factor; therefore, periodic analysis through sampling is necessary for monitoring compositional changes. Although manual sampling is typically used, it is time-consuming and can introduce errors due to low reproducibility. To address this issue, we have developed an automatic molten salt sampling device using the cold-finger method. This method involves immersing the tip of a tungsten rod in hightemperature LiCl-KCl, removing it after a few seconds, and allowing the adhered molten salt to solidify instantly. A collector then scratches and drops the solidified sample. These processes are carried out automatically using servo motors, enabling the sampling device to move around the molten salt system. We have optimized the sampling conditions, such as insertion and withdrawal rate, immersion time, and the interval between continuous sampling, based on the molten salt temperature. The temperature was set between 500°C and 850°C, considering the operating temperatures of the applications. In addition to sampling speed, the sampling depth is a key condition for determining the sampling mass. Therefore, we examined the amount of sample depending on the sampling depth and, particularly, considered the change in salt height when sampling is performed continuously. As a result, we determined the number of sampling iterations required to reach the target sample mass. Furthermore, to minimize the initial salt loss, we noted that sampling from the salt surface resulted in less representative samples. To determine the reliability, we compared the results of surface sampling with those obtained when sampling at the middle of the salt. This study will enable highly reproducible and reliable sampling by providing a prototype for an automatic sampling device for molten salt along with guidelines.

Aluminum’s exceptional properties, such as its high strength-to-weight ratio, excellent thermal conductivity, corrosion resistance, and low neutron absorption cross-section, make it an ideal material for diverse nuclear industry applications, including aluminum plating for the building envelope of nuclear power plants. However, plating aluminum presents challenges due to its high reactivity with oxygen and moisture, thus, complicating the process in the absence of controlled environments. Plating under an inert atmosphere is often used as an alternative. However, maintaining an inert atmosphere can be expensive and presents an economic challenge. To address these challenges, an innovative approach is introduced by using deep eutectic solvents (DES) as a substitute for traditional aqueous electrolytes due to the high solubility of metal salts, and wide electrochemical window. In addition, DESs offer the benefits of low toxicity, low flammability, and environmentally friendly, which makes DESs candidates for industrial-scale applications. In this study, we employed an AlCl3-Urea DES as the electrolyte and investigated its potential for producing aluminum coatings on copper substrates under controlled conditions, for example, current density, deposition duration, and temperature. A decane protective layer, non-polar molecular, has been used to shield the AlCl3-Urea electrolyte from the air during the electrodeposition process. The electrodeposition was successful after being left in the air for two weeks. This study presents a promising and innovative approach to optimizing aluminum electrodeposition using deep eutectic solvents, with potential applications in various areas of the nuclear industry, including fuel cladding, waste encapsulation, and radiation shielding.

Radioactive carbon dioxide (14CO2) capture using innovative materials is desirable due to associated radiological hazards, and growing climate change. Mineral carbonation technology (MCT) is amenable to irreversibly capture CO2. Typically, MCT is attractive because capturing carbon through the chemical reaction between alkaline earth metal ions and CO2 forms insoluble and significantly stable carbonates. However, most applications of MCT have an intrinsic restriction regarding their operational conditions since no forward reaction occurs within realistic time scales. Thereby, the CO2 capture performance, such as CO2 capacity and carbonation reaction rate, of MCTs and their applications are severely restricted by the difficulty of operations under mild conditions. For example, natural minerals require aggressive carbonation reaction conditions e.g. high pressure (≥ 20 bar), high temperature (> 373 K), and pH-adjusted carrier solutions. To overcome such obstacles, the fabrication of alkaline earth oxides impregnated into an amorphous glass structure have been recently developed. They show enhanced rates of dissolution of alkaline earth metal ions and carbonation reaction due to the loosely packed glass structure and the generation of a surface coating silica gel, consequently facilitating CO2 capture under mild conditions. In this presentation, we report the synthesis and application of a crystallized glass tailored by controlled heat treatment for CO2 capture under mild conditions. The controlled heat treatment of an alkaline earth oxide-containing glass gives rise to a structural transformation from amorphous to crystalline. The structural characterizations and CO2 capture performance, including CO2 capacity, carbonation reaction rate, and the dissolution rate of alkaline earth metal ion, were analyzed to reveal the impact of controlled heat treatment and phase transformation.

Metakaolin-based geopolymers have shown promise as suitable candidates for 14C immobilization and final disposal. It has been shown that the physicochemical properties of metakaolin wasteforms meet, and often far exceeding, the strict compression strength and leaching acceptance criteria of the South Korea radioactive waste disposal site. However, it is not possible to analyze and characterize the internal structure of the geopolymer wasteform by conventional characterization techniques such as microscopy without destruction of the wasteform; an impractical solution for inspecting wasteforms destined for final disposal. Internal inspection is important for ensuring wastes are homogenously mixed throughout the wasteform and that the wasteform itself does not pose any significant defects that may have formed either during formulation and curing or as a result of testing prior to final disposal. X-ray Computed Tomography (XCT) enables Non-Destructive Evaluation (NDE) of objects, such as final wasteforms, allowing for both their internal and external, characterization without destruction. However, for accurate quantification of an objects dimensions the spatial resolution (length and volume measures) must be know to a high degree of precision and accuracy. This often requires extensive knowledge of the equipment being used, its precise set-up, maintenance and calibration, as well as expert operation to yield the best results. A spatial resolution target consists of manufactured defects of uniformed dimensions and geometries which can be measured to a high degree of accuracy. Implementing the use of a spatial resolution target, the dimensions of which are known and certified independently, would allow for rapid dimensional calibration of XCT systems for the purpose of object metrology. However, for a spatial resolution target to be practical it should be made of the same material as the intended specimen, or at least exhibit comparable X-ray attenuation. In this study, attempts have been made to manufacture spatial resolution targets using geopolymer, silica glass, and alumina rods, as well as 3D printed materials with varying degrees of success. The metakaolin was activated by an alkaline activator KOH to from a geopolymer paste that was moulded into a cylinder (Diameter approx. 25 mm). The solidified geopolymer cylinder as well as both the silica glass rod and alumina rod (Diameter approx. 25 mm) we cut to approximately 4 mm ± 0.5 mm height with additional end caps cut measuring 17.5 mm ± 2.5 mm height. All parts were then polished to a high finish and visually inspected for their suitability as spatial resolution targets.

Chemical environments of near-field (Engineered barrier and surrounded host rock) can influence performance of a deep geological repository. The chemical environments of near-field change as time evolves eventually reaching a steady state. During the construction of a deep geological repository, O2 will be introduced to the deep geological repository. The O2 can cause corrosion of Cu canisters, and it is important predicting remaining O2 concentration in the near-field. The remaining O2 concentration in the near field can be governed by the following two reactions: oxidation of Cu(I) from oxidation of Cu and oxidation of pyrite in bentonite and backfill materials. These oxidation reactions (Cu(I) and pyrite oxidation) can influence the performance of the deep geological repository in two ways; the first way is consuming oxidizing agents (O2) and the second way is the changing pH in the near-field and ultimately influencing on the mass transport rate of radionuclides from spent nuclear fuel (failure of canisters) to out of the engineered barrier. Hence, it is very important to know the evolution of chemical environments of near-field by the oxidation of pyrite and Cu. However, the oxidation kinetics of pyrite and Cu are different in the order of 1E7 which means the overall kinetics cannot be fully considered in the deep geological repository. Therefore, it is important to develop a simplified Cu and pyrite oxidation kinetics model based on deep geological repository conditions. Herein, eight oxidation reactions for the chemical species Cu(I) were considered to extract a simplified kinetic equation. Also, a simplified kinetics equation was used for pyrite oxidation. For future analysis, simplified chemical reactions should be combined with a Multiphysics Cu corrosion model to predict the overall lifetime of Cu canisters.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Concrete is one of the largest wastes, by volume, generated during the decommissioning of nuclear facilities, which significantly influences the projected costs for the disposal of decommissioning wastes. Concrete consists of aggregates and a cement binder. In radioactive concrete, the radioisotopes are mainly associated with the cement component. If the radioactive isotope can be separated from the concrete to below the clearance criteria, the volume of radioactive concrete waste could be reduced effectively. We were studied to separate the radioactive materials from the concrete by using the thermomechanical and chemical treatment processes, sequentially. From the study, separated aggregate could be treated to achieve the clearance level. However, these processes generate a large volume of secondary acidic radioactive wastewater, which might be a critical problem to reduce the volume of radioactive concrete waste. In this research, separating the 137Cs and 90Sr from dissolved concrete wastewater to below the discharge criteria by precipitation method, it would be released to the environment under industrial waste guidelines. The experiments were conducted to using a simulated radioactive wastewater, formed by the dissolution of concrete within HCl, which was spiking the 137Cs and 90Sr, respectively. In addition, we applied the chemical precipitation methods with wastewater, using ferrocyanide for 137Cs and BaSO4 coprecipitation for 90Sr. As a result, targeted radionuclides could be removed to the discharge level (137Cs: 0.05 Bq·ml−1, 90Sr: 0.02 Bq·ml−1) by precipitation method. Therefore, it could reduce the secondary wastewater effectively by precipitation method and enhance the additional volume reduction for radioactive concrete waste.

The nuclear legacy that remains in the United Kingdom (UK) is complex and diverse. Consisting of legacy ponds and silos, redundant reprocessing plants, research facilities, and non-standard or one-off reactor designs, the clean-up of this legacy is under the stewardship of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, the NDA has made great strides in dealing with the UK’s nuclear legacy. Fuel debris and sludge removal from the legacy ponds and silos situated at Sellafield, as part of a prompt decommissioning strategy for the site, has enabled intolerable risks to be brought under control. Reactor defueling and waste retrievals across the Magnox fleet is enabling their transition to a period of care and maintenance; accelerated through the adopted ‘Lead and Learn’ approach. Bespoke decommissioning methods implemented by the NDA have also enabled the relevant site licence companies to tackle non-standard reactor designs and one-off wastes. Such approaches have potential to influence and shape nuclear decommissioning decision making activities globally, including in Korea.

The challenges facing companies and institutions surrounding civil nuclear decommissioning are diverse and many, none more so than those faced in the United Kingdom. The UK’s Generation I nuclear power plants and early research facilities have left a ‘Nuclear Legacy’ which is in urgent need of management and clean-up. Sellafield is quite possibly the most illfamed nuclear site in the UK. This complex and challenging site houses much of what is left from the early days of nuclear research in the UK, including early nuclear reactors (Windscale Piles, Calder Hall, and the Windscale Advanced Gas Cooled Reactor) and the UK’s early nuclear weapons programme. Such a legacy now requires careful management and planning to safely deal with it. This task falls on the shoulders of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, key developments in R&D, and the implementation of site licenced companies to enact decommissioning activities, the NDA aims to safety, and in a timely manner, deal with the UK’s nuclear legacy. Such approaches have the potential to influence and shape other such approaches to nuclear decommissioning activities globally, including in Korea.

본 연구에서는 우라늄 폐촉매 처리 공정에서 발생하는 우라늄 함유 폐기물 대상으로 유리-세라믹 매질 구조의 대형 디스크 소결체 형태로 제작 시, 최종 제작된 소결체의 비등방향 수축 특성 및 변형율 변화를 연구하였다. 본 연구에서는 최대 직경 40 cm를 갖는 다양한 크기 원형 디스크 형태와 원형 디스크의 1/4 크기의 부채꼴 형 소결체를 제작하여 이들의 비등방성 수축 특성을 평가하였다. 60 MPa 압력하에서 만들어지는 성형체는 소결 시 성형체의 크기 및 형태에 관계없이 높은 등방성 수축하였다. 제조된 전체 소결체에 대한 비등방성율은 평균 1.6%이었고 이때 평균 부피 감용율은 37.4% 이었다. 이러한 결과 로부터 국내에서 발생한 우라늄 폐촉매를 처리하기 위한 공정에서 발생하는 우라늄 함유 폐기물은 대형 디스크 형태의 유리-세라믹 매질 형태로 고형화함으로써 높은 안정성과 부피감용 효과를 가지며 200 L 드럼에 포장될 수 있음이 확인되었다.