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

Currently, Hanul NPP packages glass fiber classified as particulate waste in plastic packaging bags and stores them in 200 L drums. KORAD’s Waste Acceptance Criteria (WAC) presents that very low-level soil can be immobilized by loading it in a soft bag and then packaging it in a 200 L or 320 L steel drum. As currently accepted method of packaging with soft bag applies to only very low-level soils among the wastes with a risk of dispersion, it is necessary to develop a non-dispersible treatment suitable for the characteristics of other particulate waste in the future. Therefore, in order for Hanul packaging pack to be approved as an alternative method for immobilization of dispersible substances, it is necessary to verify the suitability of the packaging bag. In this paper, whether the glass fiber packaging bag used in Hanul NPP satisfies the characteristic of the soft bag presented in the WAC and the possibility of being considered as a non-dispersible measure for particulate are examined. The soft bag must meet the following requirements: material and structure, shape, drop test, and immersion test. The results of the review are as follows. First, since the glass fiber is already packaged in the drum, only the role of the inner layer, made of polyethylene, having a watertight function may be required. Second, when packaging a drum, the packaging bag is compressed into a shaped frame having an inner size of a 200 L drum, so it is packaged with little empty space in the drum. Third, as a result of a drop test of a packaging pack containing 20 kg of contents from a height of 1.2 m, it was confirmed that there was no leakage of contents. Fourth, the packaging bag was immersed in a 1-m depth water tank for 30-minutes, and the performance corresponding to the IPX7 was satisfied. As a result of reviewing the soft bag characteristic of Hanul glass fiber packaging bag, it is considered that the bag can be used as one of the non-dispersible measures because it meets almost the characteristics required by the WAC. In addition, the acceptance criteria of overseas disposal sites present various secure packaging methods in place of immobilization as a non-dispersible measure for waste containing particulate matter. It is necessary to reflect these overseas cases in the establishment of non-dispersible measures for domestic waste acceptance in the future.

In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

Colloid migration is an important topic in post-closure safety assessment of radioactive waste repository as radionuclide can be adsorbed onto colloidal particles and migrated along with the colloids. This would reduce retardation of radionuclide migration, thus increasing the released concentration into biosphere. Recently, glass fiber waste has been found to contain small sized crushed glass fiber particles (GFPs), and concerns regarding the colloidal impact of GFP is being discussed. In this study, relevance of assessing GFPs facilitated radionuclide transport in the disposal environment of 1st phase disposal facility. Colloidal impact assessment can be divided into two sections, colloid mobility, and colloid sorption assessments. Considering GFP being denser than water, fluid velocity of 1st phase disposal facility is too slow to initiate movement of such dense particles. GFPs would remain settled, and no colloidal impact is expected. In this study, sorption assessment mainly focused to analyze the possible impact if migration of GFP does occur. The GFP is mainly composed of SiO2 and few other metal oxides. Due to high composition of SiO2 in the GFPs, negative surface charge is induced onto the surface of the GFPs in alkaline environment. This negatively charged surface can attract free positive ions (ex. Ni, Co, Fe, etc.) in the repository, and these ions would be adsorbed onto the surface of the GFPs via coulomb force. Thus, if GFPs migrate, colloid facilitated radionuclide transport can be expected. However, before being released into the biosphere, particles must pass through the engineered and natural barriers, where ion-colloid-rock interactions could result in transfer of radionuclide from one media to another. At Naka Research Center, Japan, ion-colloid-rock interactions are experimented with bentonite colloid, and the result showed that despite colloid’s sorption ability was 10 times higher than the barrier material, the overall released radionuclide concentration has negligible change. To reflect such phenomenon, coulomb attractive force of GFPs and concrete is calculated and compared, which the result showed that glass fiber was 10 times weaker than concrete. Considering the Japan’s experimental result, glass fiber facilitated transport would not enhance the radionuclide release into the biosphere. Nonetheless, assuming GFPs being mobile in 1st phase disposal facility, GFPs’ sorption ability is found to be negligible compared to the concrete of the repository, thus radionuclide transport is not expected to be enhanced. In future, this study could be used as basis for further colloidal impact analysis for the safety assessment of the repository.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

Korea Radioactive Waste Agency (KORAD), regulatory body and civic groups are calling for an infrastructure system that can more systematically and safely manage data on the results of radioactive waste sampling and nuclide analysis in accordance with radioactive waste disposal standards. To solve this problem, a study has been conducted on the analysis of the nuclide pattern of radioactive waste on the nuclide data contained in low-and intermediate-level radioactive waste. This paper will explain the optimal repackaged algorithm for reducing radioactive waste based on previous research results. The optimal repackaged algorithm for radioactive waste reduction is comprised based on nuclide pattern association indicators, classification by nuclide level of small-packaged waste, and nuclide concentration. Optimization simulation is carried out in the order of deriving nuclide concentration by small-packaged, normalizing drum minimization as a function of purpose, normalizing constraints, and optimization. Two scenarios were applied to the simulation. In Scenario 1 (generating facilities and repackaged by medium classification without optimization), it was assumed that there are 886 low-level drums and 52 very low-level drums. In Scenario 2 (generating facilities and repackaged by medium classification with optimization), 708 and 230 drums were assigned to the low-level and very low-level drums, respectively. As a result of the simulation, when repackaged in consideration of the nuclide concentration and constraints according to the generating facility cluster & middle classification by small package (Scenario 2) the low-level drum had the effect of reducing 178 drums from the baseline value of 886 drums to 708 drums. It was found that the reduced packages were moved to the very low-level drum. The system that manages the full life-cycle of radioactive waste can be operated effectively only when the function of predicting or tracking the occurrence of radioactive waste drums from the source of radioactive waste to the disposal site is secured. If the main factors affecting the concentration and pattern of nuclides are systematically managed through these systems, the system will be used as a useful tool for policy decisions that can prevent human error and drastically reduce the generation of disposable drums.

Organic complexing agents which are contained in the radioactive waste can form the complex with radionuclides and enhance the solubility of radionuclides. The mobility of radionuclides to the far-field from the repository will be increased by radionuclide-ligand complex formation. Therefore, the assessment of the radionuclides’ solubility should be performed in the presence of organic complexing agents. In this study, five radionuclides (cobalt, strontium, iodine, cesium, and uranium) and three organic complexing agents (ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA)) were selected as model radionuclides and organic complexing agents, respectively. For simulating the in-situ condition, the groundwater near the repository was collected and applied in solubility experiments and the solubility was measured in various environmental conditions such as different pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and a range of organic complexing agent concentrations (10-5, 10-4, 10-3, and 10-2 M). In cases of cesium and iodine, they were very soluble in all conditions, and the effect on their solubilities was not observed. However, at high pHs, cobalt and strontium showed lower solubilities than at neutral pH and the solubility enhancement by the organic complexing agents was significant. Moreover, the effects of each organic ligand showed obvious differences and were in the order of EDTA > NTA > ISA. The solubility of uranium was increased with increasing the organic ligand concentration at lower pHs, but the organic complexing agents did not cause a remarkable difference at high pHs. According to these results, the presence of complexing agents could enhance the radionuclides’ solubility and increase the potential to release the radionuclides to the far-field from the repository. Solubility experiments of other major radionuclides in the repository are in progress.

The liquid radioactive waste system of nuclear power plants treats radioactive contaminated wastes generated during the Anticipated Operational Occurrence (AOO) and normal operation using filters, ion exchange resins, centrifuges, etc. When the contaminated waste liquid is transferred to an ion exchanger filled with cation exchange resin and anion exchange resin, nuclides such as Co and Cs are removed and purified. The lifespan and replacement time of the ion exchange resin are determined by performing a performance test on the sample collected from the rear end of the ion exchanger, and waste ion exchange resin is periodically generated in nuclear power plants. In the general industry, most waste resins at the end of their lifespan are incinerated in accordance with related laws, but waste resins generated from nuclear power plants are disposed of by clearance or stored in a HIC (High Integrity Container). Plasma torch melting technology can reduce the volume of waste by using high-temperature heat (about 1,600 degrees) generated from the torch due to an electric arc phenomenon such as lightning, and secure stability suitable for disposal. Plasma torch melting technology will be used to check thermal decomposition, melting, exhaust gas characteristics, and volume reduction at high temperatures, and to ensure disposal safety. Through this research, it is expected that the stable treatment and disposal of waste resins generated from nuclear power plants will be possible.

The “shadow zone” is defined as a region below a flow obstacle, such as a vault, in unsaturated soils. Due to the capillary discontinuity of the cavity, water saturation on the top and side of the cavity is higher than the ambient saturation. On the bottom of the cavity, however, there is a region where water saturation is lower than ambient saturation. Undoubtedly, a shadow zone may also exist below a LILW disposal vault built in subsurface soils above the water table before the vault is fully degraded. During the degradation, flow in the shadow zone is controlled by the rate of water infiltrating the degrading vault. In this study, as one of the efforts to be made for enhancing safety margin by a realistic safety assessment of the engineered vault type LILW disposal facility, the shadow zone effect is investigated by a numerical parametric study using AMBER code. The conceptual model and data were excerpted from IAEA, ISAM Vault Test Case for the liquid release design scenario. It is assumed that the nearfield barriers degrade with time. In order to compare a visible shadow zone effect, the vault degradation period is assumed to be both 500 and 1,000 years, and the shadow zone depth to be varied according to unsaturated zone lithology. It can be seen that with a shorter shadow zone (2.7 m), radionuclides arrive at the water table earlier than with a full shadow zone (55 m) due to increased advection rate in the unsaturated zone. This effect tends to be more visible in the case of a longer degradation period. For radionuclides with short residence time relative to their half-lives in the unsaturated zone, such as Tc-99 and I-129, the radionuclides are shown to come out because they will arrive sooner, thereby allowing less peak release rate, when the shadow zone effect is considered. Once the vault is completely degraded and the infiltration rate of water flowing through the vault is equal to the ambient rate, the shadow zone effect disappears. In this example calculations using IAEA ISAM Vault Test Case input parameters, it might not be shown a significant shadow zone effect. Nevertheless, when the extent of the shadow zone is determined through more sophisticated hydraulic studies in the unsaturated soils surrounding the vault, the shadow zone effect would be checked up on the realistic near-field radionuclide transport modeling in order to contribute to gaining safety margins for post-closure safety assessment of the Wolsong 2nd phase LILW disposal facility.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.

Cellulose-based wastes can be degraded into short-chain organic acids at the cementitious radioactive waste repository. Isosaccharinic acid (ISA), one of the main degradation products, can form the chelate complex with metals and radionuclides, and these complexes have a potential that can accelerate to move the radionuclides to far-field from the repository. This study characterized the amount of generated ISA from typical cellulosic materials in the repository. Two different degradation experiments were conducted under alkaline conditions (saturated with Ca(OH)2 at pH 12.4): i) cellulosic material mixture under an opened condition (partially aerobic), and ii) cellulosic material under an anaerobic condition in a nitrogen-purged glove box. In the first case, three different types of cellulosic materials–paper, cotton, and wood– were mixed at the same ratio, and the experiments were carried out at three different temperatures (20°C, 40°C, and 60°C). It revealed that both the cellulose degradation rate and generated ISA concentration were high at high reaction temperatures, and various soluble degradation products such as formic acid and lactic acid were generated. The cellulose degradation in this work seems to still stay at a peeling-off process. In the second study, each type of cellulosic material was applied in its own batch experiments, and the amount of generated ISA was in the order of paper > wood > cotton. The above two experiments are supposed to be a long-term study until the generated ISA reaches an equilibrium state.

The Nuclear Cycle Experiment Research Center is one of the facility of the Korea Atomic Energy Research Institute (KAERI). This facility is a laboratory-scale version of pyro-processing technology. Mixture depleted Uranium (DU) and depleted Uranium (DU) feed material are used in this facility for pyro-research. During summer, air conditioners that maintain temperature and humidity are always in operation to protect analysis equipments. 15 air conditioners are installed in this facility. The condensate which is generated in 15 air conditioners is collected in one place to analyze. Sampling was performed to check the level of contamination, U, pH and gamma radiation test were performed. This paper shows the degree of contamination of air conditioner condensate which is generated in the radiation management area.

Radioactive waste containing cellulosic materials such as cotton, paper and wood are being disposed in Low-and intermediate-level radioactive waste disposal site in Gyeongju. Cellulose has recently emerged great issue in terms of disposal site safety as it can be decomposed into an organic complex compound, ISA (isosaccharinic acid), under strong alkali conditions (pH 12.5 or higher) formed by the hydrated cement, to accelerate the mobility of the radionuclides in the disposal facility. However, in Korea, there are insufficient criteria for confirming the suitability for disposal of low-and intermediatelevel radioactive wastes including cellulose, and there is no specific method for evaluating the total amount of waste to confirm the suitability of disposal. Therefore, the method of SKB (Swedish Nuclear and Fuel Management Company), which has established acceptance criteria related to the physicalchemistry safety of cellulose, is analyzed to suggest a method for deriving the amount of cellulosecontaining waste disposal. Cellulose, an organic complexing agent, is an important consideration for safety case at the Swedish low-and intermediate-level radioactive waste disposal site SFR. SKB calculated the amount of cellulose generated by separately labeling cellulose-containing wastes of 1-2BMA, Silo and 1BTF (SKB 2013). BLA, a low-level radioactive waste disposal facility, is not considered due to its low radionuclide inventory (~0.2% of SFR’s total radionuclide inventory, SKB 2013). To calculate the amount of cellulose that can be disposed of, information on the mass and volume of hydrated cement (concrete waste, cement solidification waste, disposal container, grouting, disposal shed), the concentration of ISA absorbed in the hydrated cement, and the concentration of ISA dissolved in the groundwater which were used. In addition, the total disposable amount was calculated using the cellulose degradation rate, composition ratio, and the cellulose containing waste volume.

It is necessary to prepare for cutting and storing waste materials in the reactor vessel internals (RVI) for successful decommissioning of the nuclear power plant (NPP). Since RVI contain massive components and relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. To achieve efficient decommissioning waste management, this study presents radiation level of RVI and cutting optimization was performed for intermediate level waste. As a result of the radiation evaluation, a part of the core side and the upper part of RVI were evaluated as intermediate-level waste, and other components were evaluated as very low-level or lowlevel waste. For intermediate-level waste cutting, the minimum cutting method that can be put into a container was reviewed in consideration of the size, thickness, and cutting method of the interior product. The final segmentation parts are expected to be loaded into two storage containers.

Radionuclides can be leached into groundwater or soil over a long period of time due to unexpected situations even after being permanently disposed of in a repository. Therefore, it is necessary to investigate the mobility of radionuclides for the safety assessment of radioactive waste disposal. In this study, the effects of organic complexing agents such as ethylenediaminetetraacetic acid (EDTA) and isosaccharinic acid (ISA) on the sorption behavior of 239Pu and 99Tc over cementitious (concrete and grout) and natural rock samples (granite and sedimentary rock) were investigated in batch sorption experiments. For characterization of rock samples, XRD, XRF, FT-IR, FE-SEM, BET, and Zeta-potential analyses were performed. For the evaluation of mobility, the distribution coefficient (Kd) was selected and compared. The adsorption experiment was carried out at two pHs (7 and 13), a temperature of 20°C, and a range of organic complexing agents concentrations (10-7~10-2 M and 10- 5~10-2 M for 239Pu and 99Tc, respectively). The radionuclides concentrations in adsorption samples were analyzed using ICP-MS. The Kd values for 239Pu in all rock samples reduced significantly due to the presence of EDTA, even at low concentrations such as 10-5 M. In the case of ISA, the limiting noeffect concentration was much higher than that of EDTA. On the other hand, 99Tc showed relatively lower Kd values than 239Pu, and the sorption behavior of 99Tc was almost unaffected by the organic complexing agents for all rock samples. Therefore, it is possible to assume that the increased mobility of radionuclides, especially, 239Pu, in groundwater caused by the lowering of sorption at even low concentrations of organic complexing agents may result in the transport of radionuclides to the nearand far-field location of the repository.

Radionuclides stored in a radioactive waste repository over a long period of time might be leached through the barriers such as engineered rock (cement) and natural rock (granite). Organic complexing agents such as ethylenediaminetetraacetic acid (EDTA) and isosaccharinic acid (ISA) may also influence the mobility of radionuclides. In this study, a continuous fixed-column reactor packed with engineered and natural rocks was designed to investigate the effect of organic complexing agents on cesium mobility through cement and granite under anaerobic conditions. The influent flow rate of the mixed solution (organic complexing agent and cesium) at the column bottom was 0.1 mL/min, while that of groundwater was 0.2 mL/min, which was introduced between cement and granite layers in the middle of the column. The hydraulic properties such as diffusion coefficient and retardation factor were derived by a bromide tracer test. The effects of different operating parameters, such as initial cesium concentrations, initial EDTA or ISA concentrations, and bed size, on the cesium adsorption were investigated. The Thomas, Adams-Bohart, and Yoon-Nelson models were applied to the experimental data to predict the breakthrough curves using non-linear regression. These results suggest that organic complexing agents such as EDTA and ISA significantly influence the mobility of cesium in the barriers, indicating that the presence of complexing agents enhances the migration of cesium to the geosphere.

Engineered barriers (concrete and grout) in Low- and Intermediate-Level Waste (L/ILW) disposal facilities tend to degrade by groundwater or rainfall water over a long period of time. During the degradation process, radionuclides stored in the disposal facility might be released into the pore water, which can pass through the natural rock barriers (granite and sedimentary rock) and may reach the near-field and far-field. In this transportation, radionuclide might be sorbed onto the engineered and natural rock barriers. In addition, the organic complexing agent such as ethylenediaminetetraacetic acid (EDTA) and α-isosaccharinic acid (ISA), is also present in pore water, which may affect the sorption and mobility of radionuclide. In this study, the sorption and mobility of 90Sr under different conditions such as two pHs (7 and 13), different initial concentrations of organic complexing agents (from 10-5 M to 10-2 M), and solutions (groundwater, pore water, and rainfall water) were investigated in a batch system. The groundwater was collected at the L/ILW disposal facility located at Gyeongju in South Korea. The pore water and rainfall water were artificially made in the laboratory. The concrete, grout, granite, and sedimentary rock samples were collected from the same study sites from where the groundwater was collected. The rock samples were crushed to 53-150 micrometers and were characterized by XRD, XRF, SEM-EDS, BET, and zeta potential analyzer. 90Sr concentration was determined using liquid scintillation counting. The sorption of 90Sr was described by distribution coefficients (Kd) and sorption reduction factor (SRF). In the case of EDTA, the Kd values of 90Sr remained constant from 10-5 M to 10-3 M and tended to decrease at 10-2 M, while in case of ISA the Kd values decreased steadily as the concentration of ISA was increased from 10-5 M to 10-3 M; However, a sudden reduction in the Kd values were observed above 10-2 M. In comparison to EDTA, ISA gave a higher SRF of 90Sr. Therefore, from the above results, it can be concluded that the presence of ISA has a greater effect on the sorption and mobility of radionuclide in the solutions than EDTA, and the radionuclide may reach near- and far-field of the L/ILW disposal facility.

Low- and intermediate-level radioactive wastes have been disposed of in the first-phase deep underground silo disposal at Gyeongju in South Korea. These radioactive wastes contain harmful radionuclides such as Uranium-238 (238U), which can pose long-term and deleterious effects on humans and the natural environment. Ethylenediaminetetraacetic acid and isosaccharinic acid, which can be formed via cellulosic waste degradation under high alkaline conditions might considerably enhance the transport behavior of 238U with the intrusion of rainwater and groundwater. In this study, the engineered barriers (concrete and grout) and natural barriers (sedimentary rock and granite) were used to investigate the 238U transport behavior in artificial cementitious porewater of State I (pH 13.3) and State II (pH 12.5) based on groundwater or rainwater. The surface properties and geochemical compositions of barrier samples were characterized using XRD, XRF, SEM-EDX, and BET. The transport behaviors of 238U in various solution conditions were observed by sorption distribution coefficient (Kd) at a range of initial chelating agents concentration (10-5-10-2 M). The sorption behavior of 238U was retarded more in the engineered rock barriers than in the natural rock barriers. The mobility enhancement of 238U was more significant in State I than in State II. In comparison with the absence of chelating agents, negligible changes in the Kd values of 238U were observed at less than initial chelating agent concentrations of 10-4 M. However, the Kd values of 238U were significantly reduced at initial chelating agent concentrations higher than 10-3 M. Therefore, these experimental findings show that the transport behavior of 238U into the geo- and bio-sphere could be accelerated by the presence of chelating agents and the type of cement degradation states.