Chelating agents in low and intermediate radioactive wastes can form complexes with radionuclides and increase the mobility of the radionuclides. According to the Korea Radioactive Waste Agency (Acceptance criteria for low and intermediate radioactive waste, WAC-SIL-2022-1), if the amount of residual chelating agents in the waste are greater than 0.1%, the chemical names and residual amounts should be specified; if greater than 1%, the waste must be solidified and contain no more than 8%. The existing method for analyzing chelates in radioactive waste was based on UV–Visible spectrophotometry (UV-Vis), but the new method is based on liquid chromatography/mass spectrometry (LC-MS). The analysis was performed in aqueous solution before applying to real samples. Since the real sample may contain several heavy metals, it is expected that the chelates will exist as complexes. Therefore, 1.0×10-4 mol L-1 of EDTA (Ethylenediaminetetraacetic acid), DTPA (Diethylenetriaminepentaacetic acid), NTA (Nitrilotriacetic acid), and excess metals in aqueous solution were analyzed using HPLC using RP (Reverse Phase) column and HILIC (Hydrophilic interaction) column. When the RP column was used, each substance eluted without separation at the beginning of the analysis. However, when analyzed using a HILIC column, the peaks of each substance were separated. LC-MS measurements using HILIC conditions resulted in separations with better sensitivity.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of organic liquid radwaste and radiation levels are also varied. At KAERI, the organic liquid radwaste has been stored at Radioactive Waste Treatment Facility (RWTF) temporarily due to the absence of the recognized treatment technique while inorganic liquid radwaste can be treated by evaporation, bituminization, and solar evaporation process. The organic liquid radioactive waste such as spent oil, cutting oil, acetone, ethanol, etc. was generated from the nuclear facilities at KAERI. Among the organic liquid radioactive wastes, spent oil is particularly significant. According to the nuclear safety act, radioactive waste can be cleared by incineration and landfilling if it meets the criteria of less than 10 μSv/h for individual dose and 1 person – Sv/y for collective dose. Dose assessment was performed on some organic liquid radioactive waste with a very low possibility of radioactive contamination stored in RWTF at KAERI. As a result, it was confirmed that some wastes met the regulatory clearance standards. Based on this, it was approved by the regulatory body, and this became the first case in Korea and KAERI for permission for regulatory clearance of organic liquid radioactive waste by landfill after incineration.

Radioactive liquid waste generated during the operation of domestic nuclear power plants is treated through a somewhat different liquid radwaste system (LRS) for each plant. Prior to the introduction of standard nuclear power plants, LRS used a concentrated water dry system (CWDS) to evaporate liquid waste and manage it in the form of dry powder. The boron-containing radioactive liquid waste dry powder was solidified using paraffin from 1995 to 2010, and about 3,650 drums (based on 200 L) of paraffin solidified drums are currently stored in nuclear power plants. Paraffin solidification drums do not meet the acceptance criteria for radioactive waste repositories because it is difficult to secure the homogeneity of the solidified body and there are concerns about leaching of radioactive waste due to the low melting point of paraffin. In order to solve this problem and safely permanently dispose of paraffin solidification drums, the characteristics of dry powder paraffin solidification drums containing boron-containing radioactive liquid waste must be analyzed and appropriate treatment technology utilizing the results must be introduced. This study analyzes the physical properties of paraffin, the chemical properties of boron-containing radioactive waste dry powder, and the physicochemical properties of paraffin solidification powder, and proposes an appropriate alternative technology for treating boron-containing radioactive waste dry drum. When disposing of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder, the solidification body must be effectively withdrawn from the drum and the paraffin must be completely separated from the solidification body. When disposing the drum, the solidified material must be effectively extracted from the drum and the paraffin must be completely separated from the solidified material. Afterwards, the paraffin must be self-disposed, and the radioactive waste must be disposed of in accordance with acceptance criteria of repository. We looked at how each characteristic of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder can be utilized in each of the above treatment processes.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF) or on-site. In this regard, there is no Code and Standard for the amount of liquid radioactive waste generated during NPP are dismantled, but ANSI/NS-55.6 describes the amount of liquid radioactive waste generated from a light water reactor type NPP. This code is applied to nuclear power-related facilities such as domestic NPP and radioactive waste disposal facility. Therefore, this review intends to suggest an application plan for domestic NPP decommissioning through codes for liquid radioactive waste expected to generate during nuclear power plant decommissioning.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

A large amount of small and medium-sized metal waste is generated during the decommissioning of nuclear power plants (NPPs). Metal waste is mostly contaminated with low-level radioactive, so it needs decontamination for self-disposal and recycling. A large amount of Organic Decontamination Liquid Waste during decontamination will be generated. The generated organic liquid waste is low in concentration, so the decomposition efficiency is low in the decomposition process. A conditioning process is necessary to concentrate at a high concentration. For effective treatment for Organic Decontamination Liquid Waste, the composition of organic liquid waste and conditioning process were analyzed. Organic acids, metal ions, radioactive nuclides, surfactants, etc. are present in the Organic Decontamination Liquid Waste, and suspended solids are sometimes generated by various reactions. According to previous studies, the concentration of organic acids including surfactants obtained results from several tens of ppm to a maximum of 1,000 ppm, so the maximum value of 1,000 ppm was assumed. For the composition and total amount of metal ions, the average value (52.7wt% Fe, 16.3wt% Ni, 15.1wt% Cr, 15.9wt% Mn) of the distribution of metal species removed by the actual decontamination process is applied, and the total amount is 1,000 ppm was assumed. As for the radionuclides, only 60Co and 137Cs, which are expected to be mainly present, were considered, and 60Co was assumed to be 2,000 Bq/g and 137Cs to be 360 Bq/g by referring to the literature. The amounts of suspended solids were assumed to be 500 ppm by referring to the characteristics of the liquid waste generated in the decontamination process of the NPPs. Based on the estimated value, a reaction formula was established and a simulated Organic Decontamination Liquid Waste was prepared. As a result of measurement using an analysis device, the composition of the estimated and simulated Organic Decontamination Liquid Waste had similar values. The conditioning and treatment process largely consists of pretreatment, conditioning, decomposition processes. Organic Decontamination Liquid Waste goes through a pretreatment process to remove impurities with large particles. In the conditioning process, treated water that has passed through the UF/RO membrane system is discharged into the environment. At this time, Concentrated water goes through a decomposition process for processing the Organic Decontamination Liquid Waste, and is discharged to the environment through a secondary RO membrane system. The conditioning process is the low-concentration Organic Decontamination Liquid Waste in the UF membrane system is forming a micelles in an RO membrane system, concentrating it to a high concentration and then go through a recirculation process in the UF membrane system. An experiment was conducted to confirm whether the concentration of surfactants occurred during the conditioning process. As a result of the experiment confirmed that the highly concentrated surfactant formed micelles and was filtered out in the UF membrane system.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.

In KHNP CRI, the PTMs (plasma torch melting system) was developed as a treatment technology of a wide variety of radioactive wastes generated by nuclear power plants. The facility is made of melting zone, thermal decomposition zone, melt discharge zone, waste feeding device, MMI, and offgas treatment system. In this study, demonstration test was conducted using NaOH solution as liquid waste to evaluation the applicability of the PTM system. For demonstration test of NaOH solution treatment, the plasma melting zone is sufficiently pre-heated by the plasma torch for 5 hours. The temperature inside the plasma melting zone is about 1,600°C. The NaOH solution as simulant was put into the thermal decomposition zone by the spray feeding device with the throughput of maximum 30 liter/hour. During the test, the power of plasma torch is about 100 kW on the transferred mode. The 160 liters of liquid waste was treated for 500 minutes. After the demonstration test, the final product in the form of salt was remained in the melting zone, and the disposal of the final product are still under consideration.

Organic waste generated by small and medium-sized (S&M-sized) metal decontamination in NPP decommissioning. To lower the concentration of these organic substances for a level acceptable at the disposal site, the project of “Development of Treatment Process of Organic Decontamination Liquid Wastes from Decommissioning of Nuclear Power Plants” is being carried out. The conditioning and treatment process of organic liquid waste was designed. Also, the literature was investigated to make simulated organic liquid waste, and the composition of these waste was analyzed and compared. As the decontamination agent, organic acids such as EDTA, oxalic acid, citric acid are used. The sum of the concentrations of these organic materials was set to a maximum value of 1,000 ppm. The major metal ions of the decontamination liquid waste estimated are 59Fe, 51Cr, 54Mn, 63Ni, and the concentrations are respectively 527, 163, 161, 159 ppm. Additional major metal ions are 60Co, 58Co, 137Cs. 58Co is replaced by 60Co because it has the same chemical properties as 60Co. Unlike the HLW, the contamination level of S&M-sized metal in primary system was quite low, so 60Co is set to 2,000 Bq/g. Considering the contribution of fission and gamma ray dose constant, 137Cs was estimated to 360 Bq/g. Also, suspended solids of decontamination liquid waste were set at 500 ppm. Under these assumptions, the simulated organic liquid waste was made, and then organic substances and metal ions were analyzed with TOC analyzer and ICP-OES. The TOC analysis value was expected to 392 ppm in consideration of the equivalent organic quantity. the test result was 302 ppm. Some of organics appears to have been decomposed by acid. The values of metal ions (Fe3+, Cr3+, Mn2+, Ni2+) analyzed by ICP-OES are 139, 4, 152, 158 ppm, respectively. A large amount of Cr3+ and Fe3+ were expected to exist as ions, but they existed in the form of suspended solid. Mn2+ and Ni2+ came out similar to the expected values. The designed conditioning and treatment process is largely divided into pretreatment, conditioning, and decomposition processes. After collecting in the primary liquid waste storage tank, large particulate impurities and suspensions are removed through a pretreatment process. In the conditioning process, treated liquid waste passes through UF/RO membrane system, and pure water is discharged to the environment after monitoring. Concentrated water is decomposed in the electrochemical catalyst decomposition process, then this water secondarily passes through the RO membrane system and then discharged to the environment after monitoring. Through an additional experiment, the conditioning and treatment process will be verified.

Starting with the permanent shutdown of Kori Unit 1, the first waste treatment facility in Korea will be built on the Kori site. In this facility, major process such as decontamination, cutting, radiation measurement and volume reduction of decommissioning waste are performed, and radioactive liquid waste is generated by the waste treatment process and personnel decontamination. The generated liquid waste is finally discharged to the sea through radioactive monitoring system after sufficient treatment to meet the standard radiological effluent control. Whereas the treated liquid waste is additionally diluted through the circulation water discharge conduit and discharged to the sea in the operating nuclear power plants, there is no circulation water in the waste treatment facility. Therefore, a new discharging method for dilution after treatment should be considered. In this paper, the treatment concept and discharge method of radioactive liquid waste system in waste treatment facility are reviewed.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

Liquid scintillation cocktail is liquid waste, which consists of an organic solvent, scintillator, surfactant, and radionuclide. Large volumes of liquid scintillation waste are generated each year, and both the organic compound and radionuclide content can negatively affect on the health and the environment. Therefore, the liquid scintillation waste should be treated in an appropriate way. In this study, to facilitate the treatment of liquid scintillation waste, the sulfate-radical advanced oxidation process (SR-AOP) was performed for the mineralization of liquid scintillator waste. In SR-AOP, highly reactive sulfate radicals, which react more selectively and efficiently with organic compounds, are produced in situ by cleaving the peroxide bond in the persulfate molecule. For the experiment, 100 times diluted ULTIMA GOLD-LLT (initial TOC=699,800 ppm) was used as a liquid scintillation waste. The TOC removal efficiency of liquid scintillation waste by the OXONE (potassium peroxymonosulfate, PMS, 2KHSO5+KHSO4+K2SO4) and sodium persulfate (PS) with varying dosages (4–12 mM) was tested, and the effects of Co2+ and Cu2+ catalysts were compared at a range of pHs (3, 7, and 9). The experimental results demonstrated that 91% TOC removal of ULTIMA GOLD-LLT could be achieved for SR-AOP at initial pH=9, Co2+=1.2 mM (catalyst), PMS=4.8 mM (oxidant) for 60 min reaction. Compared to traditional Fenton AOP which is effective only at low pH, PMS based SR-AOP with Co2+ catalyst is effective at wide range of pHs and less dependent on the treatment efficiency of the operational pH. Therefore, it can be useful for the mineralization of liquid scintillation waste which is difficult to treat with a general treatment method due to the mixture of various organic compounds.

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.