Various types of radioactive liquid and solid wastes are generated during the operation and decommissioning of nuclear power plants. To remove radionuclides Co-60, Cs-137 etc. from a liquid waste, the ion-exchange process based on organic resins has been commonly used for the operation of nuclear facilities. Due to the considerations for the final disposal of process endproduct, other treatment methods such as adsorption, precipitation using some inorganic materials have been suggested to prepare for large amounts of waste during decommissioning. This study evaluated sintering characteristics for radioactive precipitates generated during the liquid waste treatment process. The volume reduction efficiency and compressive strength of sintered pellets were the major parameters for the evaluation. Major components of a simulated precipitate were some coagulated (oxy) hydroxides containing light elements, such as Si, Al, Mg, Ca, and zeolite particles. Green pellets compressed to around 100 MPa were heated at a range of 750~850°C to synthesize sintered pellets. It was observed that the volume reduction percentages were higher than 50% in the appropriate sintering conditions. The volume reduction was caused by the reduction of void space between particles, which is an evidence of partial glassification and ceramization of the precipitates. This result can also be attributed to conversion reactions of zeolite particles into other minerals. The compressive strength ranged from 6 to 19 MPa. These results also showed a significant correlation with the volume reduction of sintered body. Although our lab-scale experiments showed many benefits of sintering for the precipitates, optimized conditions are needed for large-scale practical applications. Evaluation of sintering characteristics as a function of pellet size and further testing will be conducted in the future.

Various dry active wastes (DAWs) have been accumulated in nuclear power plants since the DAWs are mostly combustible. KAERI has developed a thermochemical treatment process for the used decontamination paper as an operational waste to substitute for incineration process and to decontaminate radionuclides from the DAWs. The thermochemical process is composed of thermal decomposition in a closed vessel, chlorination of carbonated DAWs, separation of soluble chlorides captured in water by hydroxide precipitation, and immobilization of the precipitate. This study examined the third and fourth steps in the process to immobilize Co-60 by fabricating a stable wasteform. Precipitation behaviors were investigated in the chloride solution by adding 10 M KOH. It was shown that the precipitates were composed of Mg(OH)2 and Al(OH)3. Then, the glass-ceramic wasteform for the precipitates were produced by adding additive mixtures in which silica and boron oxide were blended with various ratios. The wasteform was evaluated in terms of volume reduction ratio, bulk density, compressive strength, and leachability.

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.

Dry active wastes (DAWs) are a type of combustible radioactive solid waste, which includes decontamination paper, protective clothing, filters, plastic bags, etc. generated from operating nuclear facilities and decommissioning projects. The volume of DAWs could be increased over time, disadvantage to higher disposal costs and space utilitization of disposal site. Additionally, incineration methods cannot be applied to DAWs, unlike general environmental waste, due to concerns about air pollution and the release of harmful chemicals with radioactive nuclides into the atmosphere. Recently, KAERI developed an alternative thermochemical process for reducing the volume of DAW, which involves a step-wise approach, including carbonization, chlorination, and solidification. The purpose of this process is to selectively separate the radioactive nuclides from carbonized DAWs that are less than clearance criteria, which can be disposed of as non-radioactive waste. In this research, we investigated the thermal decomposition characteristics of DAWs using nonisothermal thermogravimetric analysis, which was performed with different categorized wastes and heating conditions. As a result, the cellulose DAWs such as decontamination paper and cotton were thermally decomposed in three or four-step depending on the heating conditions. On the other hand, the hydrocarbon and rubber DAWs such as plastic bags and latex were thermally decomposed in one or two-step. Therefore, it could be suggested the thermochemical treatment conditions that minimize the decomposition of DAWs by controlling the reaction steps, and we will try to apply these results for cellulose type DAWs such as decontamination paper and cotton, which is generated majorly from the nuclear facilities in the future.

This study evaluated the synthesis of optimal materials for high efficiency adsorption and removal characteristics of Cs-137 for radioactive contaminated water, and considered thermal treatment methods to stabilize the spent adsorbent generated after treatment. We synthesized a composite adsorbent with a combination of impregnating metal ferrocyanide that improves the selectivity of Cs adsorption with zeolite capable of removing Cs as a support. The Cs removal efficiency of the composite adsorbent was evaluated, and the stability change of Cs according to the high-temperature sintering was evaluated as a stabilization method of the spent adsorbent. The metal ferrocyanide content of the adsorbent was in the range of 11.8~36.0%. The adsorption experiments were performed using a simulated liquid waste to have a total Cs concentration of 1 mg/L while containing a trace amount of Cs-137, and then gamma radioactivity was analyzed. In order to evaluate the stabilization of the spent adsorbent, heat treatment was performed in the range of 500~1,100°C, and the volatilization rate of Cs during heat treatment and the leaching rate of Cs after heat treatment were compared. In the adsorption experiment, the Cs removal efficiency was higher than 99%, regardless of the amount of metal ferrocyanide in the composite adsorbent. In the sintering experiment on the spent adsorbent, it was confirmed that there was no volatilization of Cs up to 850°C, and then the volatilization rate increased as the heating temperature increased. On the other hand, the leaching rate of Cs in the sintered adsorbent tends to significantly decrease as the heating temperature increases, so that Cs can be stabilized in the sintered body. In addition, as the content of metal ferrocyanide increases, the volatilization rate of Cs rapidly increases, indicating that the unstable metal ferrocyanide in the adsorbent may adversely affect the removal of Cs as well as the thermal treatment stability.

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.

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

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

As the design life of nuclear power plants are coming to the end, starting with Kori unit 1, nuclear power related organizations have been actively conducted research on the treatment of nuclear power plant decommissioning waste. In this study, among various types of radioactive waste, stabilization and volume reduction experiments were conducted on radioactive contaminated soil waste. Korea has no experience in decommissioning nuclear power plants, but a large amount of radioactively contaminated soil waste was generated during the decommissioning of the KAERI research reactor (TRIGA Mark- II) and the uranium conversion facility. This case shows the possibility of generating radioactive soil waste from nuclear power plants and nuclear-related facilities sites. Soil waste should be solidified, because its fluidity and dispersibility wastes specified in the notification of the Korea Nuclear Safety and Security Commission. In addition, the solidified waste forms should have sufficient mechanical strength and water resistance. Numerous minerals in the soil are components that can make glass and ceramics, for this reason, glass-ceramic sintered body can be made by appropriate heat and pressure. The sintering conditions of soil were optimized, in order to make better economical and more stable sintered body, some additives (such as additives for glass were mixed) with the soil and sintering experiments were conducted. Uncontaminated natural soil was collected and used for the experiment after air drying. Moisture content, pH, bulk density, and organic content were measured to understand the basic properties of soil, and physicochemical properties of the soil were identified by XRD, XRF, TG, and SEM-EDS analysis. In order to understand the distribution by particle size of the soil, it was divided into Sand (0.05–2 mm) and Fines (< 0.05 mm). The green body was manufactured in the form of a cylinder with a diameter of 13mm and a height of about 10mm. Appropriate pressure (> 150 MPa) was applied to the soil to make a green body, and appropriate heat (> 800°C) was applied to the sintered body to make a sintered body. The sintering was conducted in a muffle furnace in air conditions. The volume reduction and compressive strength of the sintered body for each condition were evaluated.

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.

Radioactive carbon, C-14, can be generated by the neutron capture reaction of O-17 during the nuclear power plant operation. Since C-14 is classified as an intermediate level waste radionuclide, it is required that an effective separation process for C-14. C-14 is mainly absorbed on activated carbon in the air cleanup system. Therefore, the main generation source of C-14 during the nuclear power plant decommissioning is spent activated carbon. KAERI has been developing the treatment of spent activated carbon. In this process, C-14 can be desorbed as a gaseous oxide form from the spent activated carbon at high-temperature vacuum conditions. This radioactive carbon dioxide can be captured into alkaline earth metal incorporated glass and can be transformed into carbonate form. However, the carbonate (e.g. CaCO3 and SrCO3) is dispersive. When the radioactive carbonates are disposed into a geological repository, they should be immobilized to remove future uncertainty. This study examined the stabilization/immobilization of the radioactive carbonates by the cement hydration process. Cement wasteform incorporated with calcium carbonate and strontium carbonate was produced under various waste loading (e.g. 20wt%, 40wt%, and 60wt% of CaCO3 and SrCO3, respectively). Then we evaluated mechanical and chemical durability by measuring compressive strength and leachability according to standard test methods specified in the waste acceptance criteria of the Gyeongju low and intermediate level waste repository (WAC-SIL-2022-1). Also, microstructure and thermal characteristics were investigated by SEM-EDS and TGA analysis.

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.