The development of existing radioactive waste (RI waste) management technologies has been limited to processing techniques for volume reduction. However, this approach has limitations as it does not address issues that compromise the safety of RI waste management, such as the leakage of radioactive liquid, radiation exposure, fire hazards, and off-gas generation. RI waste comes in various forms of radioactive contamination levels, and the sources of waste generation are not fixed, making it challenging to apply conventional decommissioning and disposal techniques from nuclear power plants. This necessitates the development of new disposal facilities suitable for domestic use. Various methods have been considered for the solidification of RI waste, including cement solidification, paraffin solidification, and polymer solidification. Among these, the polymer solidification method is currently regarded as the most suitable material for RI waste immobilization, aiming to overcome the limitations of cement and paraffin solidification methods. Therefore, in this study, a conceptual design for a solidification system using polymer solidification was developed. Taking into account industrial applicability and process costs, a solidification system using epoxy resin was designed. The developed solidification system consists of a pre-treatment system (fine crush), solidification system, cladding system, and packing system. Each process is automated to enhance safety by minimizing user exposure to radioactive waste. The cladding system was designed to minimize defects in the solidified material. Based on the proposed conceptual design in this paper, we plan to proceed with the specific design phase and manufacture performance testing equipment based on the basic design.

The sustainability of the nuclear power industry hinges increasingly on the safe, long-term disposal of radioactive waste. Despite significant innovations and advancements in nuclear fuel and reactor design, the quest for a permanent solution to handle accumulating radioactive waste has received comparatively less attention. Conventionally, two widely recognized solidification methods, namely cementation for low and intermediate-level waste and vitrification for high-level waste, have been favored due to their simplicity, affordability, and availability. Recently, geopolymers have emerged as an appealing alternative, gaining attention for their minimal carbon footprint, robust chemical and mechanical properties, cost-effectiveness, and capacity to immobilize a broad spectrum of radionuclides, including radioactive organic compounds. This study delves into the synthesis of metakaolin-based geopolymers tailored for the immobilization of fission products like cesium (Cs) and molybdenum (Mo). The investigation unfolded in two key steps. In the initial step, we optimized the alkali content to prevent the occurrence of efflorescence, a potential issue. Remarkably, as the Na2O/Al2O3 ratio increased from 0.82 to 1.54, we observed significant enhancements in both compressive strength (11.45 to 27.07 MPa) and density (up to 2.23 g/cm3). This suggests the importance of careful adjustment in achieving the desired geopolymer characteristics. The second phase involved the incorporation of 2wt% of Cs and Mo, both individually and as a mixture, into the geopolymer matrix. We prepared the GP paste, which was poured into cylindrical molds and cured at 60°C for one week. To scrutinize the crystallinity, phase purity, and bonding type of the developed matrix, we employed XRD and FTIR techniques. Additionally, we conducted standard compressive strength tests (following ASTM C39/C39M-17b) to assess the stacking durability and robustness of the developed waste form, vital for storage, handling, transportation, and disposal in a deep geological repository. Furthermore, to evaluate the chemical durability, diffusivity and leaching of the GP waste matrix, we employed the ASTM standard Product Consistency Test (PCT: C 1285-02) and American nuclear society’s devised leaching test (ANS 16.1). It is noteworthy that the introduction of Cs and Cs/Mo in the GP matrix led to a reduction of more than 50% and 60% in compressive strength, respectively. This outcome may be attributed to the interference of Cs and Mo with the geopolymerization process, potentially causing the formation of new phases. However, it is crucial to emphasize that both developed matrices exhibited an acceptable normalized leaching rate of less than 10-5 g·m-2·d-1. This finding underscores the promising potential of the GP matrix for the immobilization of cationic and anionic radioactive species, paving the way for more sustainable nuclear waste management practices.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

Wet solid wastes including spent ion exchange resins, evaporator concentrates and sludges require solidification to transform wastes into an acceptable solid, monolithic form for final disposal. The development of the process control program for the solidification of radioactive sludges generated at nuclear power plants has been in progress to provide reasonable assurance that the solidified product will meet the established waste acceptance criteria for solidified waste. A mobile solidification system to produce the solidified waste in the size of a 200 L drum was used, which adopts the in-line mixing method where the waste and binder are mixed and then transferred to the disposable container. To simulate radioactive sludges, non-radioactive sludges are synthesized and the specimens are prepared by using them. The qualification tests on the prepared specimens including the compressive strength test, the thermal cycling test, the irradiation test, the leach test, the immersion test, etc. have been performed to qualify recipes for a range of waste compositions. The results of the tests will be analyzed and discussed.

The soils contaminated with radionuclides such as Cs-137 and Sr-90 should be solidified using a binder matrix, because radioactively contaminated soils pose environmental concerns and human health problems. Ordinary Portland cement has been widely used to solidify various radioactive wastes due to its low cost and simple process. In this study, simulant soil waste was solidified using cement waste form. The soils were collected around ‘Kori Nuclear Power Plant Unit 1’ and they were contaminated with the prepared simulant liquid waste containing Fe, Cr, Cs, Ni, Co, and Mn. The water-to-dry ingredients (W/D) ratio of cement waste form was 0.40. The cement paste was poured into a cubic mold (5×5×5 cm) and then cured for 28 days at room temperature. The 28-day compressive strength, water immersion, and EPA1311-toxicity characteristic leaching procedure (TCLP) tests were performed to evaluate the structural stability of cement waste form. The compressive strength was not proportional to soil waste loading, and the lowest compressive strength (4±0.1 MPa) was achieved in cement waste form containing 50wt% soil waste. After the water immersion test for 90 days, the compressive strength of cement waste form with 50wt% soil waste increased to 7.5±0.6 MPa, meeting the waste form acceptance criteria in the repository. It is believed that long-term water immersion test contributed to the additional curing and hydration reaction, resulting in the enhanced compressive strength. As a result of the TCLP test, the released amount of As, Ba, Cd, Cr, Pb, Se, Co, Cs, and Sr was less than the domestic and international standards. These results imply that cement waste form can be a promising candidate for the solidification of radioactive soil wastes.

A lot of solid wastes are generated when nuclear power plant is dismantled, and a lot of treatment costs and optimal waste treatment technologies are required to treat the generated solid wastes. Currently, there is no optimized reduction and solidification technology for each characteristics of radioactive dismantling waste, so the customized treatment technology for each waste is required to respond actively to this issue. This paper shows the evaluation results of molding and sintering characteristics using preliminary sample to derive operational characteristics and improvements for powder mixing device, molding device, and sintering device manufactured for solidification of dispersible radioactive waste. Zeolite was used as a preliminary sample for performing basic operation characteristics evaluation of each unit device. First of all, the basic operation characteristics of the powder mixing device was evaluated by analyzing the sample distribution, mixing degree, and tap density. It was confirmed that the preliminary sample was well mixed in all areas of the cylinder where the mixing was performed. In the tap density analysis, the increase effect of the volume reduction of the sample was confirmed according to the increase of the RPM speed (up to 2000 RPM). Since the particle size of zeolite sample is very small (nanometer size), the particular consistency of the change of average particle size with RPM speed couldn’t be confirmed, but the uniform of particle size distribution was confirmed with RPM speed size. The basic operation characteristics of the molding device was evaluated for each mold size (ID30, ID50, ID100) according to the moisture content (0-20%) and the molding pressure condition (25-200 MPa) for the preliminary sample. In the characteristics evaluation of the sintered body, the strength of the sintered body was much higher than that of the molded body. However, it was confirmed that as moisture evaporated during the sintering process according to the moisture content contained in the molded body, the swelling occurred in the sintered body due to vapor pressure, and this caused cracks in the longitudinal or transverse direction inside and outside the sintered body. Therefore, optimal moisture content conditions for sintering should be derived. In conclusion, if the operation characteristics and improvements of powder mixing, molding and sintering devices derived from this study are reflected and improved, it is judged that it is possible to derive the optimal process for solidification of dispersive radioactive wastes.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

Present study investigated the waste form integrity of melted products generated from PAM-MSO system, which is proposed and developed to compensate the drawbacks of each system. The disposal suitability of the melting solidification products generated from the plasma arc melting treatment of pulverized cement debris spiked by Pb, Cd and Cs, as indicators of typical hazardous metals and radionuclides existed in the low-level mixed waste in the KHNPPs. The final waste form obtained by the test was evaluated for suitability for disposal. The compressive strength was 261.10 MPa, showing much higher values when compared to other waste form products. The compressive strength of both the sample after irradiation with 107 Gy radiation and that after long-term submersion test (90 days) satisfied the disposal criteria. As a result of the leaching test conducted according to the ANS 16.1 test method, it was confirmed that the leaching index satisfies the disposal criteria.

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.

Encapsulation using cement as a solidification method for disposal of radioactive waste is most commonly used in most advanced countries in the nuclear technology to date due to its advantages such as low material cost and accumulated technology. However, in case of cement solidification, since moisture or hydroxyl group in cement is decomposed by radioactivity, it may happen that gas is generated, structural stability is weakened, and leachability is increased due to low chemical durability. Therefore, the various new solidification methods are being developed to replace it. As one of these alternative technologies, for dispersible metal compounds generated through the incineration replacement process, the study on engineering element technology for powder metallurgy is under way, which overcomes the interference problem between mechanical elements and media that may occur during the process such as the homogeneous mixing process of the target powder substance and additives used in the powder metallurgy concept-based sintering process for the solidification of the final glass composite material (GCM), the process of creating a compressed molded body using a specific mold, the process of final sintering treatment. The solidification process of dispersible radioactive waste can be largely divided into pre-treatment stage, molding stage, and sintering stage, and the characteristics of the final radioactive waste solidification material can vary depending on the solidification treatment characteristics of each stage. In relation with these characteristics, the matters to be considered when designing device for each stage to solidify dispersible radioactive waste (property of super-mixing device for homogenized powder formation, structural geometry and pressure condition of molding device for production of compressed molded body, temperature and operation characteristics of sintering device for final glass composite material (GCM), etc.) are drawn out. It is expected that the solidification device design reflecting these considerations will meet all disposal conditions of radioactive waste material, such as compressive strength and leaching characteristics of solidified radioactive waste material, and create a uniformized solidification of radioactive waste material.

KHNP-CRI has developed Mega-Watt Class PTM (Plasma Torch Melter) for the purpose of reducing the volume of radioactive waste and immobilizing or solidifying radioactive materials. About 1 MW PTM is a treatment technology that operates a plasma torch and puts drum-shaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. Since only the overflowing slag is discharged from the melter, the discharge is intermittent. Therefore, solidification occurs in the process of discharging the melt. It is difficult to accumulate evenly in the waste container, and there is also an empty space. Solid radioactive waste must be disposed of to meet the acceptance criteria for radioactive waste. Plasma-treated solid waste raised concerns about the requirements. The waste solidification output in a slag container gave us some concerns for the waste package’s solidification and encapsulation requirements. The plasma-treated solid waste process to meet the acceptance criteria will be cost and need time consuming. Thus, a induction heating will be introduced to meet solidification requirements and test criteria of the solidification waste for the waste package disposal.

The number of nuclear power plants that are permanently shut down or decommissioned is increasing worldwide, and accordingly, research is being conducted on an appropriate method for disposing of radioactive waste generated during the decommissioning of nuclear power plants. In the case of waste liquid generated during the decommissioning of nuclear power plants, it is important not only to efficiently reduce waste but also to secure the suitability of disposal. One of the solidification treatment methods for radioactive waste is cement solidification, but since cement solidification has poor solidification properties and generates a large amount of waste, improvement activities have been pursued. This study aims to develop high-performance cement-based materials and solidification treatment technology for solidification of liquid radioactive waste generated during nuclear decommissioning in order to improve the problems of cement solidification treatment method. For the development of polymer cement, epoxy resin and polyamine/amide mixed type and general Portland cement were mixed in various ratios. The most appropriate mixing ratio was 4.5:2, which showed the highest compressive strength. A simulated waste liquid was prepared by referring to the preliminary decommissioning plan of Shin-Kori Units 5 and 6, and it was dried and made into granules. Polymer cement was injected into a drum filled with granules by vacuum pressure to prepare a waste form matrix. In the solidification process, granules made by drying the waste liquid were used, and the solidification agent was filled in between the granules, so the total volume of solid radwaste was reduced compared to the conventional cement solidification treatment method. As a result, the amount of waste decreased to about 1/3, and the volume reduction rate increased by about 2.2 times. The compressive strength of 3,243 psi was confirmed in the disposability performance test for the manufactured solid samples. The compressive strength after the thermal cycling test, irradiation test, microorganism test, and immersion test was 2,257 psi, 2,306 psi, 4,530 psi, and 2,263 psi, respectively, exceeding the acceptance criteria of 500 psi. The leaching index was 7~13, and no free standing water was generated.

Magnesium potassium phosphate cements (MKPCs) are prepared by the acid-base reaction of dead burned magnesia (MgO) and monopotassium phosphate (KH2PO4). Low-pH cementitious materials such as MKPCs are currently of interest for the geological disposal of nuclear waste. MKPCs have advantages such as high early strength, high bonding strength, small drying shrinkage, low permeability, and high sulfate resistance. According to the results of previous studies, it is known that cesium, strontium, and cobalt are immobilized in the form of MgCsxK1−xPO4·6H2O, MgxSr1−xKPO4·6H2O, and Co3(PO4)2, respectively, in MKPCs. However, these results were predicted based on thermodynamic data, not directly observed precipitates to clearly show the evidence. Therefore, in this study, we directly analyzed the immobilized forms of Cs, Sr, and Co, respectively. CsNO3, Sr(NO3)2, and Co(NO3)2·6H2O powders (0.3 mol each) were mixed individually in each of the MKPC suspensions. The suspensions in which KH2PO4 was dissolved were pH 4.3 and the dissolution of MgO decreased the H+ concentration, raising the pH close to 11. The hydration products according to pH evolution in the MKPC suspensions were analyzed, and the change in the concentration of ions in the aqueous solution was also measured. An aqueous solution was obtained using a syringe filter (0.45 μm) to analyze the ion concentrations in the solution of the suspension. The collected solutions were diluted with nitric acid and analyzed using inductively coupled plasma mass spectrometry. To characterize the solid phases, the suspensions were obtained with a pipette at specific times and filtered under a vacuum in a Buchner funnel. Because the amounts of hydration products including Cs, Sr, and Co were small, it was not observed by XRD and TGA analysis, but their components could be analyzed by SEM-EDS. The final precipitate forms of Cs, Sr, and Co in the MKPC matrix are MgCsPO4·6H2O, SrHPO4, and Co3(PO4)2·8H2O, respectively.

The mixing powder of vitrification material and metallic oxide sludge was solidified by hot isostatic press method and was tested to check whether the solidified waste disposal acceptance criteria were met or not. From various contaminated tank in nuclear power plants, and other nuclear energy facilities, radioactive sludge based on metallic oxide can be generated. The most of tank consist of stainless steel can be oxidated by the long-term exposure on oxygen and moisture, and then can be made sludge layer based on metallic oxide on the inner wall of contaminated tank. Radioactive sludge waste should be solidified and disposed. Melting and hardening is the most basic method for solidification. The melting points of metallic oxide of stainless steel as Fe3O4, NiO, Cr2O3 are 1597, 1955, 2435, respectively. Those are very high temperature. To melt these metallic oxides, a furnace capable of raising the temperature to a very high temperature is required, which requires a lot of thermal energy, which may lead to an increase in disposal cost. Therefore, it is necessary to lower the melting point and solidify non-melted metallic oxide powder by adding vitrifying material powder as Na2O, SiO2, B2O3. The more vitrification material is added, the easier it is to solidify the sludge based on metallic powder at a low temperature, but there is a problem in that the total waste volume increases due to the addition of vitrification material. In this study, the mixing ratio and temperature conditions that can fix the sludge while adding a minimum amount of vitrification material will be confirmed. Mixing ratio conditions of the vitrification material and sludge powder are 10:90, 15:85, 20:80, 25:75. To fix the metallic oxide sludge by melting only the vitrification material without completely melting the metallic oxide, compression by external pressure is required. Therefore, the HIP (Hot Isostatic Pressing) method was used to solidify the metallic oxide sludge by simultaneously heating and pressurizing it. Because the softening points of most of vitrification based on Na2O, SiO2, B2O3 are ranged from 800 to 1000, temperature conditions are 800, 900, 1000. Since the compressive strength for disposing of the solidified materials was 3.4 MPa, the maximum pressure condition was set to 5000 psi (about 34 MPa), which is 10 times 3.4 MPa. And optimal mixing ratio, temperature, pressure conditions that meet the solidified waste disposal acceptance criteria will be found.