This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

During the initial cooling period of spent nuclear fuel, Cs-137 and Sr-90 constitute a large portion of the total decay heat. Therefore, separating cesium and strontium from spent nuclear fuel can significantly decrease decay heat and facilitate disposition. This study presents analytical technique based on the gas pressurized extraction chromatography (GPEC) system with cation exchange resin for the separation of Sr, Cs, and Ba. GPEC is a micro-scaled column chromatography system that allows for faster separation and reduction volume of elution solvent compared to conventional column chromatography by utilizing pressurized nitrogen gas. Here, we demonstrate the comparative study of the conventional column chromatography and the GPEC method. Cation exchange resin AG 50W-X12 (200~400 mesh size) was used. The sample was prepared at a 0.8 M hydrochloric acid solution and gradient elution was applied. In this case, we used the natural isotopes 88Sr, 133Cs, and 138Ba instead of radioactive isotopes for the preliminary test. Usually, cesium is difficult to measure with ICP-OES, because its wavelengths (455.531 nm and 459.320 nm) are less sensitive. So, we used ICP-MS to determine the identification and the recovery of eluate. In this study, optimized experimental conditions and analytical result including reproducibility of the recovery, total analysis time and volume of eluents will be discussed by comparing GPEC and conventional column chromatography.

Chelate resin is a resin that has an exchange group which can form chelates with various metal ions. It shows higher selectivity for metal ions than ion exchange resin and can selectively remove characteristic metal ions. In an aqueous solution containing metal ions, chelate resin can adsorb specific metal ions, and the separated chelate resin can desorb the adsorbed metal ions by changing temperature or pH, so chelate resin has the advantage of being reusable. Chelate resin has been used industrially as an adsorbent to adsorb and separate heavy metal ions in wastewater, and is also used for the purpose of recovering precious or rare metals contained in industrial wastewater or industrial waste. Against this background, there is a need to develop chelate resins with higher adsorption capacity. Acrylic fiber is defined as a man-made fiber made from a linear synthetic polymer with fiberforming ability consisting of more than 85% acrylonitrile. It is a man-made fiber that is often used as a substitute for wool because it has good thermal insulation properties like wool and is warm and soft to the touch. It is a fiber rich in cyano groups due to its high content of acrylonitrile, and has the advantage of being able to be used as a variety of functional fibers through modification of cyano groups. In this study, the amination reaction of acrylic fiber was performed using diethylenetriamine, and the adsorption characteristics for metal ions were evaluated according to the reaction conversion rate. In order to improve the amination efficiency, 400 kGy was irradiated using a 2.5 MeV electron beam accelerator, and through this, the crosslinking rate of acrylic fiber was able to be improved up to 80%. Water and ethanol were used as cosolvents for the amination reaction in a ratio of 60/40 vol/vol, respectively, and a reaction yield of 178% was obtained after 120 minutes of reaction. Using the chelate resin prepared in this way, the adsorption performance for metal ions was evaluated through Atomic Absorption Spectrometry analysis.

The process of carbonization followed by a high-temperature halogenation removal of radionuclides is a promising approach to convert low-radioactivity spent ion-exchange (IE) resins into freereleasable non-radioactive waste. The first step of this process is to convert spent ion-exchange resins into the carbon granules that are stable under high-temperature and corrosive-gas flowing conditions. This study investigated the kinetics of carbonization of cation exchange resin (CER) and the changes in structures during the course of carbonization to 1,273 K. Both of model-free and modelfitted kinetic analysis of mixed reactions occurring during the course of carbonization were first conducted based on the non-isothermal TGAs and TGA-FTIR analysis of CER to 1,272 K. The structural changes during the course of carbonization were investigated using the high-resolution FTIR and C-13 NMR of CER samples pyrolyzed to the peak temperature of each reaction steps established by the kinetic analysis. Four individual reaction steps were identified during the course of carbonization to 1,273 K. The first and the third steps were identified as the dehydration and the dissociation of the functional group of —SO3-H+ into SO2 and H2O, respectively. The second and the fourth steps were identified as the cleavage of styrene divinyl benzene copolymer and carbonization of pyrolysis product after the cleavage, respectively. The temperature and time positions of the peaks in the DTG plot are nearly identical to those of the peaks of the Gram Schmidt intensity of FTIR. The structural changes in carbonization identified by high-resolution FTIR and DTG are in agreement with those by C-13 NMR. The results of a detailed examination of the structural changes according to NMR and FTIR were in agreement with the pyrolysis gas evolution characteristics as examined by TGA-FTIR.

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.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

PURPOSES : The purpose of this study was to evaluate the performance of a titanium dioxide (TiO2) asphalt surface treatment agent for reducing NOx on the roadside at laboratory and full scales. METHODS : To verify the NOx reduction performance of TiO2 and silicon-based resin-applied surface treatment agents at the lab scale, a bed flow photo reactor test (ISO standard) and a mixed tank photo reactor test designed to apply real-scale construction materials were conducted. Subsequently, the full-scale NOx reduction performance was verified using a full-scale demonstration facility, and the field construction capability of the TiO2 asphalt surface treatment agent was verified through actual road site application. RESULTS : The bed flow photoreactor and mixed tank photoreactor methods showed the same trend in the NOx removal performance. Evaluation of the NOx removal performance of the TiO2 surface treatment agent revealed that the NO removal rate was approximately 13% at the laboratory scale and 15% at full scale. CONCLUSIONS : Through this study, it was determined that the asphalt surface treatment agent applied with TiO2 will have a sufficient NOx reduction effect in an actual road site. In the future, it will be necessary to analyze the continuity of the effect according to traffic volume through continuous monitoring in the field.

The incorporation of vertically aligned carbon nanotubes (VACNTs) between composites plies has been said to enhance the through-thickness strength, and it can also decrease the risk of interply delamination and reduce crack initiation. Thanks to these high mechanical performances, nano-engineered hybrid composites are seen as promising for highly demanding structural reinforcement applications. This paper is part of a study that focuses specifically on the methodology for transferring VACNTs onto a prepreg surface while maintaining their initial vertically aligned morphology. The chosen method involved bonding the VACNTs’ forest through capillary impregnation of the forest by the prepreg’s resin. Key parameters for an effective transfer and to achieve a partial capillary rise of the resin into the VACNTs will be discussed here.

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.

In this study, we have fabricated the phenolic resin (PR)/polyacrylonitrile (PAN) blend-derived core-sheath nanostructured carbon nanofibers (CNFs) via one-pot solution electrospinning. The obtained core-sheath nanostructured carbon nanofibers were further treated by mixed salt activation process to develop the activated porous CNFs (CNF-A). Compared to pure PAN-based CNFs, the activated PR/PAN blend with PR 20% (CNF28-A)-derived core-sheath nanostructured CNFs showed enhanced specific capacitance of ~ 223 F g− 1 under a three-electrode configuration. Besides, the assembled symmetric CNF28-A//CNF28-A device possessed a specific capacitance of 76.7 F g− 1 at a current density of 1 A g− 1 and exhibited good stability of 111% after 5,000 galvanostatic charge/discharge (GCD) cycles, which verifies the outstanding long-term cycle stability of the device. Moreover, the fabricated supercapacitor device delivered an energy density of 8.63 Wh kg− 1 at a power density of 450 W kg− 1.

This study was performed to evaluate the separation of Sr, Cs, Ba, La, Ce, and Nd using gas pressurized extraction chromatography (GPEC) with anion exchange resin for the quantitation of Neodymium. GPEC is a micro-scaled column chromatography system that provides a constant flow rate by utilizing nitrogen gas. It is overcome the disadvantages of conventional column chromatography by reducing the volume of elution solvent and shortening the analysis time. Here, we compared the conventional column chromatography and the GPEC method. The whole analysis time was decreased by nine times and radioactive wastes were reduced by five times using the GPEC system. Anion exchange resin 1-X4 (200~400 mesh size) was used. The sample was prepared at a 0.8 M nitric acid in methanol solution. The elution solvent was used at a 0.01 M nitric acid in methanol solution. Finally the eluate was analyzed by ICP-MS to determine the identification and recovery. In this case, we applied the natural isotopes of LREEs (139La, 140Ce, and 144Nd) and high activity nuclides (88Sr, 133Cs, and 138Ba) instead of radioactive isotopes for the preliminary test; as a result, unnecessary radioactive waste was not produced. The recoveries were 93.9%, 105.9%, 91.9%, 47.6%, 35.9%, and 79.9% of Sr, Cs, Ba, La, Ce, and Nd, respectively. The reproducibility of recoveries by GPEC were in the range 2.8%–10.9%.

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

The intermediate level spent resins waste generated from water purification for the the moderator and primary heat transport system during operaioin of heavy water reactor (HWR). Especially, moderator resins contain high level activity largely because of their C-14 content. So spent resins are considered as a problematirc solid waste and require special treatment to meet the waste acceptance criteria for a disposal site. Various methods have been studied for the treatment of spent resins which include thermal, destructive, and stripping methods. In the case of solidification methods, cement, bitument or organic polymers were suggested. In the 1990s, acid stripping using nitric acid and thermal treatment methods were actively investigated in Canada to remove C-14 nuclide from waste resin. In Japan, thermal distructive method was studied in the 1990s. Since 2005, KAERI developed acid stripping method using phosphate salt. However, acid stripping method are not suitable due to large amounts of 2nd waste containing acid solution with various nuclides. To solve this probelm, KAERI has been suggested the microwave treatment method for C-14 selective removal from waste resin in the 2010s. Pilot scale demonstration tests using radioactive waste resin generated from Wolsung unit 1 and unit 2 were successfully conducted and 95% of C-14 was selectively removed from the radioactive waste resin. In recent years, price of C-14 source is dramatically increased due to market growth of C-14 utilization and exclusive supply chain depending on China and Russia. High purity of C-14 were captured in HWR waste resin. Interest of C-14 recovery research from HWR waste resin is currently increased in Canada. In this study, microwave method is suggested to treat HWR waste resin with C-14 recovery process. Additionally, status of waste resin management and research trends of HWR waste resin treatment are introduced.

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.