Commercial operation of KORI Unit 1 ended in 2017, and the final decommissioning plan is currently under approval from the KINS. In order for the dismantling waste to go to the repository, it is judged that the radioactive waste generated during the commercial operation should be treated and disposed in advance. Among these radioactive wastes, spent filters contain various radionuclides. The radiation dose rate from the radiation coming out of the filters ranges from a low dose rate to high dose rate. Therefore, in order to handle the spent filters, a remote processing system is required to reduce the radiation exposure of workers. This paper evaluates the radioactive inventory of filters that are stored in the filter room at the KORI unit #1. For this purpose, a method for predicting the radioactivity of each nuclide in the filter, based on the radiation dose rate, has been described using the MicroShield code, which is a commercial shielding code. The information on the filters in the field has only the creation date, type, size, and surface dose rate. In order to evaluate the radioactivity inventory using such limited data, it is possible to know the nuclide radioactivity ratio in the filter. We took out some of the filters stored on site and measured from using the ISCOS system, a gamma nuclide analyzer. The radioactivity of each nuclide in the filter was inferred by modeling with the MicroShield code, based on the radiation dose rate and the radioactivity value of each nuclide measured in the field.

KHNP-CRI has developed small-capacity and 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 drumshaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. The small-capacity PTM is a treatment technology that operates a plasma torch and puts small amounts of radioactive waste by directly putting it into the melter through a waste input machine. Mega-Watt Class PTM was able to inject radioactive waste in drums, so it was disposed of without backloging. On the other hand, The small-capacity PTM put radioactive waste without a package, and the waste input was blocked. If even small-capacity PTM put radioactive waste in the form of small packages such as drums, it is expected that various types of radioactive waste can be processed for a long time. Packaging also reduces the risk of radioactive contamination.

Concrete radioactive waste is divided into surface-contaminated concrete and activated concrete, and although the generation rate varies depending on the operating conditions of the nuclear power plant, it is reported that the amount of surface-contaminated concrete generated is greater. It is reported in the ‘US-NRC Inventory Report’ that 99% of radionuclides in surface-contaminated concrete are distributed within 1 mm of the surface. Since concrete radioactive waste accounts for a large amount of generation after metal radioactive waste, it is necessary to reduce the amount of radioactive waste disposal by applying appropriate treatment techniques to surface-contaminated concrete. In this study, a similar contamination environment work space with the size of 5.4 (W) × 3.6 (L) × 2.5 (H) [m] in which concrete specimens can be fixed on the wall and floor was established. And an integrated decontamination equipment was verified the automation performance for ‘location accuracy’, ‘radioactive contamination level measurement’ and ‘concrete surface laser scabbling’. It was confirmed that the average was 8.3 [mm] in the evaluation of the ‘location accuracy’ for the remote control and movement of the integrated decontamination equipment. For performance verification of ‘radioactive contamination level measurement’ and ‘laser scabbling’, it were used that size of 30×30×8 [cm] ordinary concrete specimens and concrete radioactively contaminated with Co-60 below the regulatory exemption concentration. ‘Radioactive contamination level measurement’ is measured as much as the set range, divied and display the measured values in different colors on the map of the control program. Ordinary concrete specimens are 0.066~0.089 μ Sv/hr, and contaminated concrete specimens are 0.107~0.121 μ Sv/hr, and it was confirmed that they are expressed in different colors on the map. For ‘laser scabbling’, the performance according to the laser scabbling speed and reproducibility with ordinary concrete specimens was verified. As a result, a weight change of up to 1.48 kg was confirmed. Contaminated concrete specimens were subjected to a direct method using a surface contamination detector and an indirect method using a smear paper to measure surface contamination before and after scabbling, and the debris generated after scabbling was analyzed using HPGe.

In order to permanently dispose of radioactive waste drums generated from nuclear power plants, disposal suitability must be demonstrated and the nuclides and radioactivity contained in the waste drums, including those in the shielding drums, must be identified. At present, reliable measurements of the nuclide concentration are performed using drum nuclide analysis devices at power plants and disposal facilities during acceptance inspection. The essential functions required to perform nuclide analysis using the non-destructive assay system are the correction for self-attenuation and the dead time correction. Until now, measurements have mainly been performed for drums containing solid waste such as DAW drums using SGS calibration drums with ordinary iron drums. However, for drums containing non-uniform radioactive waste, such as waste filters embedded in cement within shielding drums, a separate calibration drum needs to be produced. In order to produce calibration drums for shielded and embedded waste drums, the design considered the placement of calibration sources, setting of shielding thickness, correction for medium density, and cement mixing ratio. Based on these considerations, three calibration drums were produced. First, a shielding drum with an empty interior was produced. Second, a density correction drum filled with cement was produced to create apparent density on the surface of the shielding drum. Third, a physical model drum was produced containing a mock waste filter and cement filled in the shielding drum.

Spent filters contained in drums of radioactive waste generated from nuclear power plants are contaminated with various radioactive isotopes due to their use in various water purification processes in the system. Radiation doses from the spent filters can vary from low to high levels. To dispose of drums containing spent filters as radioactive waste, the inventory of radioactive isotopes in the filters must be determined. Two methods for determining the inventory are indirect measurement using scaling factors and direct analysis of filter samples. This study suggests a method to determine the appropriate sample size for each drum based on the number of filters stored in the drum, when direct analysis is used to determine the inventory of radioactive isotopes. In particular, Visual Sample Plan (PNNL) software’s Item Sampling function was used to calculate the sample size, considering the confidence level and minimum acceptable coverage rate. As a result, assuming that the number of filters packed per drum ranges from a minimum of 1 to a maximum of 30, the study suggests that a full inspection is required for drums containing 9 or fewer filters, while drums containing 10 filters should be sampled with 9 samples, 11 filters with 10 samples, 12-13 filters with 11 samples, 14-16 filters with 12 samples, 19-22 filters with 14 samples, 23-26 filters with 15 samples, and 27-30 filters with 16 samples.

LILW disposal repository in Gyeongju, South Korea is considered with a concrete mixture that uses Ordinary Portland Cement (OPC) partially substituted with supplementary cementitious materials (SCMs). The degradation of cementitious materials that result from chemical and physical attacks is a major concern in the safety of radioactive waste disposal. We present a reactive transport model utilized as one of the geochemical simulation approaches for the timescales of concern that range from hundreds to thousands of years. The purpose of this study is to investigate the sensitivity of parameters in concrete disposal systems and to evaluate the influence of various assumptions on the chemical degradation of the systems using a reactive transport model. A reactive transport model in the concrete disposal vault was developed to evaluate the behavior of engineered barriers composed of cementitious materials. The sensitivity analysis was performed using reactive transport models through the coupling between COMSOL and PHREEQC. The databases selected for the analysis are the Thermochimie database presented by ANDRA. Among many variables considered, two variables that can highly affect chemical degradation were selected for detailed sensitivity analysis for dealing with uncertainties. This is important because the chemical degradation mechanism is generally sensitive to precipitation and diffusion coefficient. The first factor is precipitation, which might be the most important factor in chemical degradation because it acts as a calcium leaching of cementitious materials in a disposal system in a highly alkaline environment, increasing the porosity of the system. To predict the change in annual precipitation, the measurement of the precipitation observatory station in the nearest area of Gyeongju for the past 80 years was collected. The second factor is the diffusion coefficient, which plays an essential role in the durability of the concrete disposal system, promoting the decalcification of cementitious minerals, accelerating system degradation, and increasing the porosity of its system, thereby facilitating the migration of radionuclides. The diffusion coefficient values used in studies similar to this work were calculated and evaluated using the box-and-whisker method. The results of the sensitivity analyses for the reactive transport model in the concrete disposal system will be presented. The sensitivity cases show that the results obtained are much more sensitive to changes in transport parameters.

Decommissioning of Nuclear Power Plant (NPP) projects in South Korea starts with permanent shutdown of Kori unit 1 and Wolsung unit 1. It is important to establish a treatment and disposal method for radioactive waste generated during the decommissioning of the nuclear power plants. Large quantities of the wastes during decommissioning of NPP are generated in a short period of time and the wastes have various types and characteristics. For efficient decommissioning of NPP process, the radioactive waste is classified by types and each treatment method and packaging concept is presented respectively in this paper. Radioactive waste generated during decommissioning of NPP is classified into reactor vessel, reactor internals, metals, Dry Active Waste (DAW), concreate, spent fuel storage rack, spent resin and spent filter, etc., and the packaging concept for each type should be established to meet the waste acceptance criteria. Major waste acceptance criteria requirements include nuclides concentration, filling rate, free water, surface radiation does rate and weight. Radioactive waste containers can be classified into packaging containers, transport containers, and disposal containers. The packaging container is used to contain, transport, and store radioactive waste within the radiation control area, and a control number has been assigned as a radioactive waste drum after the final treatment has been completed. The transport container is used for transporting radioactive waste filled-containers from a radiation control area through an uncontrolled area. In this paper, the concept of disposal of dismantled radioactive waste and packaging methods were reviewed in comprehensive consideration of domestic radioactive waste transport and storage regulations, permanent disposal environment, and development status of large containers.

KHNP’s vitrification technology introduced a commercialized vitrification facility to the Hanul nuclear power site after a commercialization test through a lab test and a pilot plant at KHNP-CRI. France’s ANADEC (consortium with CEA, Orano, ECM Technologies and Andra) conducted a feasibility evaluation from FY2018 to FY2021 to apply In-Can vitrification, which was developed to treat Fukushima Effluent Treatment Waste (FETW) such as carbonate slurry and ferric slurry generated from ALPS (Advanced Liquid Processing System-Multi Radionuclides Removal) facilities for waste treatment in Fukushima, Japan. For commercialization, the following method was used. First, through the Laboratory scale studies, the possibility of high waste loading (60wt% in dry mass) of slurry on borosilicate matrix was tested. In addition, the volatility of radionuclide was evaluated through radionuclides surrogates with a Bench-scale mockup and glass discharge (100 kg) was evaluated through In-Can vitrification process verification. The feeding system was improved through a pilot scale test, and finally, glass discharge (300 kg) was evaluated after large amount of waste was treated through an industrial prototype (Fullscale) at the CEA Marcoule site (France).

As the decommissioning of nuclear power plants progresses, interest in the inevitably generated radioactive waste is also increasing. Especially, because the containers of ILW packages are significantly more expensive than the containers of LLW packages, the special attention should be focused on minimizing the number of the containers of ILW packages. The radiation dose limit for packaging of ILW shall not exceed 2 mSv/h and 0.1 mSv/h on contact and at 2 m, respectively in South Korea. Meanwhile, The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging and cutting plan, the number of containers for ILW packages generated from decommissioning of Reactor Vessel Internal (RVI) of a nuclear power plant that has been in operation for decades was optimized in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

A disposal of radioactive wastes is one of the critical issues in our society. Considering upcoming plans for dismantling of nuclear power plants, this problem is inevitable and should be discussed very carefully. There are variety of methods to handle with radioactive wastes, including Incineration, conventional gasification and plasma gasification. Among them, plasma gasification process is in the limelight due to its eco-friendly & stable operation, and large volume reduction effects. However, a fatal disadvantage is that it consumes more electric power than other methods, this leaves us a question of whether this process is indeed economical. Within the scope of this paper, I would like to introduce 4 cases which plasma facilities were evaluated economically in worldwide, and reach the conclusion on the economic feasibility of plasma process.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

Domestic nuclear power plants developed the scaling factors for the radioactive waste generated from 2004 to 2008 for the first time. Afterwards, the effectiveness of continuous application of the scaling factors have been evaluated for newly generated radioactive waste over the past two years. It was confirmed that most of the initially developed scale factors were effective within a factor of 10 principle, which is an acceptable range. The scaling factors were updated using the analysis data base from 2004 to 2016. A scaling factor refers to the calculated abundance ratio between Key (Easy-to- Measure) and DTM (Difficult-to-Measure) nuclide at the time of generation of radioactive waste based on the source term in the reactor of an operating nuclear power plant. The effectiveness of continuous application of scaling factors can be evaluated at regular intervals regardless of operation status or when there are events that change scaling factors during nuclear power plant operation, such as zinc injection, large-scale facility replacement, and long-term shutdown etc. Even in the case of a permanently shut down nuclear power plant in which fuel is withdrawn from the reactor and generation of new nuclides by nuclear fission and radiation has stopped, periodic verification is conducted to confirm whether the scaling factor developed before permanent shutdown can be effectively applied to the radioactive waste generated after permanent shutdown. However, depending on the nuclear power plant decommissioning strategy or conditions, the period of permanent shutdown prior to decommissioning can be very long, so preparations are needed to ensure the appropriateness of scaling factor operation. In the case of domestic nuclear power plants, Kori Unit 1, a light water reactor, was permanently shut down in June 2017, and as a heavy water reactor nuclear power plant, the permanent shutdown of Wolseong Unit 1 was finally decided in December 2019 after twists and turns including large-scale facility replacement and long-term shutdown. In this paper, we have predicted when the scaling factors will change significantly due to radioactive decay and the difference in halflife between the Key and DTM nuclides over time after permanent shutdown. We also have tried to find appropriate countermeasures for the operation of scaling factors during permanent shutdown period, such as updating scaling factors or applying correction factors.

Licensing for the application of the Polymer Concrete High Integrity Container (PC-HIC) to nuclear power plants has been completed or is in progress. Approval for the expanded application to all domestic nuclear power plants has been completed to utilize the 860 L PC-HICs for the 2nd stage surface repository, and the regulatory body is reviewing the license application to use the 510 L PCHICs for the 1st stage underground repository in the representative nuclear power plants. The 860 L PC-HICs, which have been licensed for all domestic nuclear power plants, will be used for safe storage management and disposal of low-dose dried concentrate waste and spent resin, and a total of 100 units is expected to be supplied to representative nuclear power plants that have been licensed first. The 510 L PC-HICs are planned to be used for underground disposal of high-dose spent resin and dried concentrate waste. Prior to the application of PC-HICs to nuclear power plants and disposal to the repository, it is necessary to establish realistic and reasonable requirements through close consultations between waste generator and disposal operators to ensure the suitability for disposal of PC-HIC packages and to carry out disposal delivery and acceptance work. Since the Polymer Concrete High Integrity Container (PC-HIC) has long-term integrity of more than 300 years and the barrier does not temporarily collapse, spent resin and dried concentrate waste, which are radioactive wastes to be solidified, can be disposed of much more safely in PC-HIC packages than solidified types. Acceptance criteria for the PC-HIC packages should be prepared fully reflecting the advantages of PC-HIC, and quality assurance methods for physical/chemical/radiological characterization results based on the Waste Certification Program (WCP) should be supported. In addition, infrastructure should be secured for safe transportation, handling, and storage of the PC-HIC packages. In this paper, we have tried to find a reasonable acceptance criteria, quality assurance method, and infrastructure level according to the dose and disposal conditions of PC-HIC packages.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

Hanford site has been operated since 1943 to produce the plutonium for nuclear weapons. Significant amount of radioactive wastes was generated by the nuclear weapons production process. The radioactive wastes are stored in 177 aged underground tanks. Due to the risk of leakage into the air and the Columbia River, the US DOE and EPA, and Washington State Department of Ecology organized the Tri-Party Agreement (TPA) to clean-up the Hanford site in 1989. The LAW (low-activity waste) vitrification facility named WTP (Waste Treatment Plant) is plan to vitrify about 212 million liters of radioactive waste. The US DOE announced that the world’s largest melter to vitrify the LAW was heated up on October 8, 2022.

After the Fukushima accident, significant amount of radioactively contaminated waste has been generated with 50~250 m3/day and stored in tanks of the Fukushima Daiichi nuclear power plant site. The contaminated water is treated by various treatment facility such as KURION, SARRY, Reverse Osmosis, and ALPS to remove 62 radioactive nuclides except H-3. For the contaminated water treatment process, massive secondary wastes such as sludge, spent adsorbent, and so on as by-product are being generated by the facilities. In Japan, to treat the secondary wastes, melting technologies such as GeoMelt, In-can vitrification and Cold Crucible Induction Melting vitrification are considered as a candidate technologies. In this study, the technologies were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for treatment of the secondary wastes.

Air conditioning facilities in nuclear power plants use pre-filters, HEPA filters, activated carbon filters, and bag filters to remove radionuclides and other harmful substances in the atmosphere. Spent filters generate more than 100 drums per year per a nuclear power plant and are stored in temporary radioactive waste storage. Plasma torch melting technology is a method that can dramatically reduce volume by burning and melting combustible, non-flammable, and mixed wastes using plasma jet heat sources of 1,600°C or higher and arc Joule heat using electric energy, which is clean energy. KHNP CRI & KPS are developing and improving waste treatment technology using MW-class plasma torch melting facilities to stably treat and reduce the volume of radioactive waste. This study aims to develop an operation process to reduce the volume of bag filter waste generated from the air conditioning system of nuclear power plants using plasma torch melting technology, and to stably treat and dispose of it. It is expected to secure stability and reduce treatment costs of regularly generated filter waste treatment, and contribute to the export of radioactive waste treatment technology by upgrading plasma torch melting technology in the future.

In this work, we report test results for direct melting of non-combustible wastes by using a 100 kW class transferred type plasma torch. For this purpose, non-combustible wastes consisting of metals and sands were prepared, weighed and melted by a transferred arc in a ceramic crucible with inner diameter of 150 mm. Test results reveal that 75wt% M6 iron bolts mixed with 25wt% sands were completely melted down within 140 seconds at the plasma power level of 83.8 kW, producing melting speed of 100 kg/hr and volume reduction rate of 62.8%. In addition, for simulated wastes consisting of 77.3wt% metal chips and 22.7wt% sands, the volume reduction rate high than 88% was achieved at 50 kW plasma power. These results indicate that non-combustible wastes can be treated efficiently when directly melting them by using transferred type plasma torch.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

Nowadays, transferred type arc plasma torches have been widely present in industrial applications, in particular, using melting pool of electrically conducting materials such as arc furnace, welding and volume reduction of radioactive wastes. In these applications, the melting pools are normally employed as an anode, thus, heat flux distributions on anode melting pool need to be characterized for optimum design of melting pool system. For this purpose, we revisited the one-dimensional model of the anode boundary layer of arcs and solved governing equations numerically by using Runge-Kutta method. In addition, the direct melting process of non-combustible wastes in the crucibles were discussed with the calculation results.