To safely dispose of highly radioactive spent resin and concentrate waste generated through nuclear power plant operations, it is essential to meet the physicochemical properties requirements of the packages and ensure the accuracy and reliability of radiological characteristics determination. Both spent resin and concentrate are packaged in high-integrity containers (HICs) after drying and are homogeneous waste products generated in the primary system and liquid radioactive waste treatment system. Meeting the physicochemical properties requirements does not appear to be difficult. However, to achieve reliable radiological characterization of high-integrity container packages, it is necessary to take a representative sample and perform accurate radiological analysis. Therefore, this paper discusses the methodology for evaluating the radionuclide inventory of high radioactive resin and concentrate packages, as well as the essential element technology and considerations. For relatively high radioactive resin and concentrate packages, the radionuclide inventory for each package should be evaluated with high reliability through direct radiological analysis of the representative samples collected for each package. This can contribute to the efficient operation of radioactive waste disposal facilities. Radionuclide-specific concentrations directly analyzed for each package will be managed in a database. As analytical data accumulates and direct measurements of high-integrity container package such as the radwaste drum assay system (RAS) become feasible, statistical techniques such as correlation analysis between easy-tomeasure (ETM) nuclides and difficult-to-measure (DTM) nuclides can lead to the development of efficient and reasonable indirect evaluation methods, such as scaling factor and the mean activity concentration method. As for the element technology, a remote representative sampling technique should be developed to safely and effectively take representative samples of highly radioactive materials, including granulated or hardened concentrate waste. Considerations should also be given to determining the sample quantity representing each package, as well as establishing radiation calibration and measurement methods appropriate to the radiation levels of the representative samples.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

Ion exchange resins are commonly employed in the treatment of liquid radioactive waste generated in nuclear power plants (NPP). The ion exchange resin used in NPP is a mixed-bed ion exchange resin known as IRN-150, which is of nuclear grade. This resin is a mixture of cation exchange resin and anion exchange resin. The cation exchange resin removes cationic radionuclides such as Cs and Co, while anion exchange resin handles anions (e.g., H14CO3 -), effectively purifying the liquid waste. Spent ion exchange resins (spent resin) containing C-14 are classified as low and intermediate level radioactive waste, and their radioactivity needs to be reduced as it exceeds the disposal limit regulated by law. Therefore, the microwave technology for the removal of C-14 from spent resin has been investigated. Previous studies have successfully developed a method for the effective removal of C-14 during the resin treatment process. However, it was observed that, in this process, functional groups in the resin were also removed, resulting in the generation of off-gases containing trimethylamine. These off-gases can dissolve in water from process, increasing its pH, which can subsequently hinder the recovery of C-14. In this study, we investigated the high-purity recovery of C-14 by adjusting the moisture content within the reactor following microwave treatment. Mock spent resins, consisting of 100 g of resin with HCO3 - ion-exchanged and 0, 25, or 50 g of deionized water, were subjected to microwave treatment for 40 or 60 minutes. Subsequently, the C-14 desorption efficiency of the mock spent resins was evaluated using an acid stripping process with H3PO4 solution. The functional group status of the mock spent resins was analyzed using 15N NMR spectroscopy. The results showed that the mock spent resins exhibited efficient C-14 recovery without significant functional group degradation. The highest C-14 desorption efficiency was achieved when 25 g of deionized water was used during microwave treatment.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

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.

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

Mixed-bed ion exchange resin consist of anion exchange resin and cation exchange resin is used to treat liquid radioactive waste in nuclear power plants. C-14 from heavy water reactors (HWR) is adsorbed on the anion exchange resin and is considered intermediate-level radioactive waste. The total amount of radioactivity of C-14 in spent ion exchange resin exceeds the activity limits for the disposal facility. Therefore, it is necessary to reduce the radioactivity through pre-treatment. There are thermal and non-thermal methods for the treatment of spent ion exchange resin. However, destructive methods have the problem of emitting off-gas containing radionuclides. To solve this challenge, various methods have been developed such as acid stripping, PLO process, activity stripping, thermal treatment and others. In this study, spent ion exchange resin (spent resin) was treated using microwave. The reaction characteristics of the resin to microwave were used to selectively remove the C-14 on the functional groups. Simulated spent anion exchange resin and spent resin from Wolseong NPP were treated with the microwave method, and the desorption rate was over 95%. An integrated process system of 1 kg/batch was built to produce operating data. After the operation of the process, characterization and evaluation of post-treatment for condensate water and adsorbent used in the process were performed. When the process system was applied to treat simulated spent resin and real spent resin, both showed a desorption rated of more than 97%. It means that the C-14 was successfully removed from the radioactive spent resin.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

The feasibility study of synthesizing graphene quantum dots from spent resin, which is used in nuclear power plants to purify the liquid radioactive waste, was conducted. Owing to radiation safety and regulatory issues, an uncontaminated ion-exchange resin, IRN150 H/OH, prior to its use in a nuclear power plant, was used as the material of experiment on synthesis of graphene quantum dots. Since the major radionuclides in spent resin are treated by thermal decomposition, prior to conducting the experiment, carbonization of ion-exchange resin was performed. The experiment on synthesis of graphene quantum dots was conducted according to the general hydrothermal/solvothermal synthesis method as follows. The carbonized ion-exchange resin was added to a solution, which is a mixture of sulfuric acid and nitric acid in ratio of 3:1, and graphene quantum dots were synthesized at 115°C for 48 hours. After synthesizing, procedure, such as purifying, filtering, evaporating were conducted to remove residual acid from the graphene quantum dots. After freeze-drying which is the last procedure, the graphene quantum dots were obtained. The obtained graphene quantum dots were characterized using atomic force microscopy (AFM), Fourier-transform infrared (FT-IR) spectroscopy and Raman spectroscopy. The AFM image demonstrates the topographic morphology of obtained graphene quantum dots, the heights of which range from 0.4 to 3 nm, corresponding to 1–4 graphene layers, and the step height is approximately 2–2.5 nm. Using FT-IR, the functional groups in obtained graphene quantum dots were detected. The stretching vibrations of hydroxyl group at 3,420 cm−1, carboxylic acid (C=O) at 1,751 cm−1, C-OH at 1,445 cm−1, and C-O at 1,054 cm−1. The identified functional groups of obtained graphene quantum dots matched the functional groups which are present if it is a graphene quantum dot. In Raman spectrum, the D and G peaks, which are the characteristics of graphene quantum dots, were detected at wavenumbers of 1,380 cm−1 and 1,580 cm−1, respectively. Thus, it was verified that the graphene quantum dots could be successfully synthesized from the ionexchange resin.