In biosphere assessment modeling for the safety assessment of the Wolsong LILW disposal facility, the multi-compartment modeling in which all radionuclides transport is described quantitatively in terms of transfer factors between various environmental compartments has been implemented. In order to reflect the actual transfer mechanisms of 14C in the environment the specific activity (SA) modeling approach can be applied as an alternative to the previous transfer factors (TF) approach. The assumption of full SA equilibrium throughout the terrestrial environment is completely satisfactory for 14C release to the atmosphere if the 12C is emitted as 14CO2. This is the only form that is readily taken up by plants, so that active carbon is incorporated into the plant via photosynthesis at the same rate as stable carbon. Accordingly, the 14C concentration in Bq/g stable carbon is the same in the plant as it is in the air. And animals take up carbon almost entirely through ingestion and the SA ratio in the plant is maintained in the animal. In this study, a specific activity model for 14C was implemented in a GoldSim biosphere assessment model. From the literature survey for existing specific activity models developed, the IAEA model was selected. The farming scenario utilizing well water was simulated and the resulting ingestion dose conversion factors (DCFs) from the IAEA SA model were compared with those of the TF approach. The parameter value for the concentration of stable carbon in the air (gC/m3) is used as 0.20 gC/m3 considering the Suess effect. The dose coefficient for food ingestion used for dose calculations was taken from ICRP-72 as 5.8E-10 Sv/Bq. It was found that the ingestion DCFs of the SA model showed about 3 times lower than those of the TF model in the farming scenario through irrigation of well water, so it is expected that the SA approach could be applied for a more realistic assessment. Though the comparisons were made on the results from the terrestrial ecosystem only in this study, it would be necessary to investigate the applicability of the SA modeling approach for 14C through extensive comparisons and analysis including an aquatic ecosystem, and through parameters survey suitable to the domestic condition.

The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

Bentonite containing >50wt% montmorillonite is being considered as a buffer material in a deep geological repository to dispose of high-level radioactive wastes (HLRW). Bentonite is considered a buffer material because of its exceptional properties such as high swelling capacity, low hydraulic conductivity, and high radionuclide sorption capacity. The bentonite buffer can be exposed to heat from the radioactive decay of HLRW and to groundwater. Water in bentonite buffer can be converted to steam under elevated temperature and pressure conditions. Previous studies reported contrasting results showing that steam treatment could decrease the swelling capacity due to changes in the surface properties from hydrophilic to hydrophobic or could not change. The contrasting results were probably because different studies used different experimental conditions and methods. Therefore, the effect of steam treatment on the bentonite properties is still unclear. The purpose of this study is to determine how the bentonite properties change after steam treatment, in particular swelling and hydrophilic properties. Two types of bentonite were used for steam treatment and analysis; Gyeongju Ca-bentonite (KJ- II) and Wyoming Na-bentonite (GCL-B). Steam treatment was performed at 150°C in an oven for various periods (7, 30, 60, and 90 days). Free swell test, X-ray fluorescence (XRF) analysis, surface-area measurement (BET), thermal gravimetric analysis (TGA), cation exchange capacity (CEC), and water uptake test were performed on steam-treated bentonite for various periods and raw bentonite. After steam treatment, some properties of steam-treated bentonite changed when compared to raw bentonite. Free swell index, which means the swelling capacity, decreased significantly as the results of previous studies. CEC and BET surface area values depended on the bentonite type. For Wyoming Na-bentonite, in which the dominant interlayer cation is a monovalent cation, CEC and BET surface area values were increased. On the other hand, Gyeongju Ca-bentonite, in which the dominant interlayer cation is a divalent cation, has no change in the above two properties. Results of XRF analysis, TGA, and water uptake test showed that these properties of both bentonites did not change after steam treatment. The results of this study confirmed that steam treatment affected the swelling and physicochemical properties of bentonite, in particular Na-bentonite. Further studies will focus on the surface properties of bentonite to investigate whether the surface properties have changed from hydrophilicity to hydrophobicity, or whether the montmorillonite structure has changed.

Excavation Damaged Zone (EDZ) is created by the excavation of deposition holes and disposal tunnels at high-level radioactive waste repository that causes macro- and micro-fracturing in the surrounding rock. Since EDZ can significantly increase the hydraulic transmissivity in the rock and act as a major pathway of leaked radionuclides, consideration of EDZ in terms of safety assessment is very important. Moreover, long-term stress changes such as stress redistribution due to excavation of nearby deposition holes and disposal tunnels, thermal stress due to temperature rise, effective stress change due to pore pressure change, and swelling pressure of bentonite buffer can increase EDZ size and change in thermal-hydraulic-mechanical properties, and consequently, it can affect the transport of radionuclides. Therefore, in order to analyze the effect of long-term evolution of EDZ on radionuclide transport, it is essential to conduct numerical analysis considering the coupled Thermal-Hydraulic- Mechanical (THM) behavior in EDZ. In order to simulate the behavior of EDZ, coupled THM model was developed using the Adaptive Process-based total system performance assessment framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). The concept of damage was introduced to demonstrate the jointed rock as a continuous medium. Among several damage models, Mazars damage model was applied in this study. Mazars damage model is the most well-known model for concrete which has similar behavior with rock as brittle material, and the input data of the model can be easily obtained through laboratory testing. If damage occurs due to the influence of thermal-hydraulic-mechanical coupled behavior at the bedrock, the properties change according to the degree of damage, and as a result, the migration of the radionuclide is affected. Based on this conceptual model, radionuclide transport model in the near field considering the long-term evolution of EDZ was developed. To investigate the effect of EDZ in terms of process-based performance assessment, the modeling results with and without EDZ were compared. Finally, by simulating the coupled THM behavior of EDZ with damage model, the effect of long-term evolution of EDZ on radionuclide transport was investigated.

Deep geological disposal is generally accepted to be the most practical approach to handling radioactive wastes. Bentonite has been considered as a buffer material in deep geological disposal repositories (DGR) for high-level radioactive wastes. Evaluating the effect of short-term bentonite alteration on EBS performance has limitations in safety assessment over thousands of years. Information on bentonite characteristics under various conditions obtained from natural systems can be used to evaluate long-term safety of bentonite buffer. The purpose of this study was to investigate mineralogical and physicochemical characteristics of bentonite in the Naah mine located in Yangnam-myeon, Gyeongju-si for a natural analogue of the bentonite barrier in DGR. A total of 15 samples were collected at regular intervals from the bentonite layer and andesitic lapilli tuff (i.e., parent rock) at the boundary with the bentonite layer. The bentonite layer is located at a depth of about 1 m below the ground surface. Each sample was separated into particles < < 75 μm and particles < 2 μm through grinding and sedimentation processes. The separated subsamples were characterized mineralogically and physiochemically using various analytic techniques. Bentonite samples have a similar SiO2/Al2O3 ratio to the parent rock and a lower (Na+K)/Si ratio than the parent rock, indicating depletion of alkali components during bentonitization. The parent rock and bentonite samples have similar mineral composition (i.e., quartz, feldspars, opal-cristobalite-tridymite and montmorillonite). Results of XRD analysis on the randomly distributed particles < 2 μm indicate that bentonite is mostly composed of Ca-montmorillonite, which is a typical dioctahedral smectite. Results of FTIR and VNIR analysis indicate that montmorillonite contained in bentonite is Al-dioctahedral montmorillonite, and Al is substituted with Mg in some octahedron units. The mineralogical and physicochemical characteristics are similar regardless of sampling location. These results suggest that bentonite potentially exposed to weathering, located near the ground surface, has hardly altered.

Backfill is one of the main components of engineered barrier in a high-level waste repository. The material selection of the backfill determines the barrier performance of the backfill. Overseas, its related research has been carried out mainly in Sweden, Finland, Canada, and Japan. However, Korea has recently started backfill research, and it is urgent to select a potential material for establishing the concept of backfill material and conducting backfill research. This study reviews NEA report, potential materials for overseas backfill research, advantages and disadvantages of single and mixed backfill materials, cases of license applications in Finland and Sweden for the selection of potential materials for backfill in Korea’s high-level waste repository. The review results indicated that it is reasonable to carry out backfill research according to the following plan: Both single and mixed materials are considered as potential materials for backfill research; experiments and performance studies are conducted with these materials; and, based on the results, a potential material or candidate material for the backfill suitable for the HLW repository in Korea is determined. For this plan, the single material is tentatively selected, as in Sweden, as bentonite with a montmorillonite content of about 40-50%. Then, if the selection criteria for montmorillonite content are determined through experiments and performance studies, we determine the final potential backfill material. As for the mixed backfill material, the bentonite/crushed rock mixture seems to be more advantageous than the bentonite/sand mixture considering the disposing problem of crushed rock generated from tunnel excavation and economic feasibility through its recycling. It is thought that the bentonite used in the bentonite/crushed rock mixture should have a higher montmorillonite content than bentonite used as a single backfill material since the crushed rock acts as an inert material in the mixture. The results of this study can be used as basic data for selecting the backfill material to be applied to the high-level waste repository in Korea, and can be used as a guideline for selecting the potential material required for backfill experiments and performance studies to be carried out in the future.

Bentonite, which mostly consists of montmorillonite, is considered as a suitable buffer material for disposal of high-level radioactive wastes in deep geological repository due to its high swelling capacity, low permeability, and strong retention capacity of radionuclide migration. Alkaline and saline solutions originated from degradation of cementitious material and seawater intrusion, respectively, may cause the changes in mineralogical and chemical properties of montmorillonite with various processes such as cation exchange within the interlayer, dissolution of montmorillonite, and precipitation of second minerals. In this study, montmorillonite alteration under alkaline and saline environments and its influences on retention of cesium and iodide by bentonite buffer were investigated. The reactions of bentonite (Bentonil-WRK) with alkaline solutions (0.1 M KOH and NaOH) and simulated saline solution were performed for 7 days in batch experiments at 25°C. After the experiments, reacted bentonite samples were characterized by X-ray diffraction (XRD), Fourier Transform Infrared (FTIR) spectroscopy, Short Wavelength Infrared (SWIR) spectrometry. The concentrations of cesium and iodide dissolved in the solutions were analyzed using an inductively coupled plasma mass spectrometer (ICP–MS). The XRD patterns showed significant decrease in the interlayer space of montmorillonite after the reaction with alkaline solution due to cation exchange and change in hydration status at the interlayer. The retention of cesium and iodide in alkaline and saline solutions were affected by montmorillonite alteration and ion competition. Therefore, the montmorillonite alteration affecting the nuclide retention capacity and long-term stability of bentonite buffer should be considered in the safety assessment of long-term geological disposal performance.

When the radioactive nuclides are leaked from a deep geological repository by groundwater, the migration path of the nuclides is mostly consisted of rock fractures to the surface biosphere. Thus, assessing the safety of the disposed radioactive wastes depends upon understanding of nuclide migration in the fractured rocks. Fractures in rocks tend to dominate the hydrological characteristics of the dissolved nuclides. To study migration of nuclides in the rock fracture, a granite block of 1 m scale was quarried from the Hwangdeung site. The block has a single natural fracture. The six faces of the rock including fracture gaps were sealed with silicone adhesives to prevent leaking or diffusion of the water. Usually flow in fractured rock is unevenly distributed and most of the water flow occures over a small portion of the fracture zone, that is so called channeling flow. It is caused by uneven distribution of apertures in a fracture field. Flow rate is proportional to the cubic of the aperture. Thus, figuring out aperture distribution in a fracture field is the most important step on the study of the migration of nuclides in the fractured region. The ideal way to figure out the aperture distribution in a fractured rock is to use a non-destructive tool such as X-ray tomagraphe. However, it has a limitation of scale, that is, less than about 30 cm. It is not easy to give a good resolution for this quarried rock of 100×60×60 cm scale. It gives complex and vague images of the fracture. The optimum way to get an aperture distribution in a fractured rock is to drill some boreholes to the fracture and to carry out hydraulic tests. The more number of boreholes gives the more accurate information, but the more disturbance to the fracture field. Thus, it is necessary to optimize between aperture information and disturbing fracture field by selecting a suitable number of boreholes. We drilled nine boreholes from the upper surface of the rock mass just to the fracture without penetrating the fracture. And we carried out dipole tests for the matrix set of 9 boreholes. From each dipole test, an effective average aperture was calculated with the data of flow rate and hydraulic head. Then aperture distribution in the fracture field is calculated with a modified Krigging method. As a result, the aperture is distributed in the range of about 0.03~0.16 mm.

In order to dispose of spent nuclear fuel (SNF) in deep geological repository, source term evaluation considering its specification, enrichment, burnup, cooling time should be performed. In this study, the measured values of Takahama-3 pressurized water reactor SNF (WH 17×17) samples were analyzed with SCALE 6.1/ORIGEN-S and TRITON code calculation results for validation. Unlike the ORIGENS code, TRITON code calculations differed from two-dimensional neutron flux distribution by using the multi-group cross-section library. Both calculation results from ORIGEN-S and TRITON code showed higher errors in 234U, 239Pu, and 241Pu compared to other actinide nuclides. In the case of axial locations of fuel rods in fuel assembly, fuel rods located at the edge of the fuel assembly presented increased errors due to nuclear reaction cross-section. Overall, the ORIGEN-S predictions informed more accurate agreement with the measured results compared with TRITON results. Especially to 235U, 239Pu, and 240Pu radionuclides, ORIGEN-S errors were denoted more than twice as low as the TRITON results. Comparing the calculation results with experimental results implied that the ORIGENS code was more accurate code than the TRITON code for source term evaluation.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

The high-level nuclear waste (HLW) repository is a 500-1,000 m deep geological disposal system with a very long life expectancy for disposing of high-level waste, which is known to have a half-life of several thousand years. This repository is subject to harsh environmental conditions, such as high temperature and radiation from high-level waste, that can cause deterioration and crack. When radiation escapes through cracks, it can injure persons on the ground. Therefore, it is essential to install a sensor that can detect problems such as cracks. But, since the high-level nuclear waste (HLW) repository is sealed with bentonite and backfill, the sensor cannot be removed or replaced once it has been installed. Therefore, it is necessary to develop a highly durable monitoring sensor that can withstand harsh environmental conditions. Before attempting to improve durability, it is first required to assess durability quantitatively. And an accelerated life test is a widely used method for assessing durability. However, it is important to obtain the same failure mode when conducting a reliability test, such as an accelerated life test. If the accelerated life test is conducted using different failure modes, the dependability of the results is inevitably diminished. Therefore, in this study, a representative failure mode for the piezoelectric sensor used in the accelerated life test was derived through experiments and literature research.

Spent nuclear fuel still emits radionuclides and high heat that are dangerous to humans. In order to permanently isolate such spent nuclear fuel from human living areas, research is underway to construct a deep disposal system (500 m underground bedrock) consisting of natural and engineering barriers. In this study, plugs, which are engineering barriers consisting of disposal containers, buffer, backfill and plugs were investigated. The plug is one of the engineered barriers made of concrete to block the outflow of radioactive materials and the ingress of organisms, through the tunnel crosssection seals that are eventually discarded. General concrete leachate has a pH of 12.5 or higher and is highly alkaline, which induces dissolution of SiO2 components contained in the buffer and backfill. Dissolved SiO2 causes precipitation and cementation on the surface of the buffer and backfill, reducing performance. Therefore, the use of low-ph concrete is essential for deep, high-level waste disposal sites. Currently, Finland, Sweden, France, Switzerland, etc. have proposed low-ph concrete mix design and performance standards. For example, in Finland, cement, silica fume and fly ash are used as binders and the compressive strength is 50 MPa or more, and the leachate pH is 11 or less. In this research, test specimen fabrication and physical property tests (strength, pH) were performed based on mix design, proposed in Finland, Sweden, France and Switzerland. A cubic (50 mm×50 mm×50 mm) and a cylinder (Ø100 mm×200 mm) specimens were fabricated. Cubic and cylinder were made of mortar and concrete, respectively, depending on whether they included coarse aggregate. General concrete strength shows the characteristic that 70 to 80% of the 28th day of the second order appears on the 14th day of the second order and converges after the 28th day. As a result of mortar strength property evaluation, it increased by 30% from 90th to 28th. pH characterization was evaluated according to the powder dissolution method (ESL method) and leaching method (Leachate, EPA 1315) on cubic (mortar) and cylindrical (concrete) specimens, respectively. Mortar ph was measured at 9.78, a decrease of up to 20% from 90 days to 7 days. The physical property evaluation of concrete is currently underway and shows a trend of increasing strength and decreasing pH according to age. Consequently, we aim to present a low-ph concrete mix design for domestic highlevel radioactive waste disposal sites.

Identifying plausible scenarios is necessary to evaluate the performance of the repository reliably over a very long period. All features, events, and processes (FEPs) expected in the repository should be comprehensively well-defined and structured into scenarios based on the relation analysis. A platform for the FEP DB management and relation analysis is needed to facilitate the efficient composition of the scenarios. For this purpose, the CYPRUS program was developed, but abandoned due to suspended FEPs and scenario research. Thus, it became necessary to build a new easy-tomaintain platform that inherits the legacy of CYPRUS and reflects the latest research. The data structure and user interface configuration were derived to develop a new platform. The new platform provides extensive data such as the assessment context, the FEP DB, the interaction between FEP contents, the relevance to other project FEPs, the influence on performance, the scenarios for the TSPA, the AMF, and the PA Data. The platform displays the long-term evolution FEPs developed by KAERI, the international and major project FEPs in table format. The correlation between FEP items is composed of a detailed interaction matrix and visualized as the chord diagram or arc diagram. The relevance and linkages between the project FEP items are mapped and presented in the form of network diagrams and network tables. The platform designed in this study will be used to manage the FEP DB, analyze and visualize the relationship between the FEP and scenarios, and finally construct the performance assessment scenarios. It is expected that the platform itself will be used as a part of the knowledge management system and facilitate efficient collaboration and knowledge exchange among experts.

Various radionuclides are released and contaminate soils by the nuclear accidents, nuclear tests and disposal of radioactive waste. Among radionuclides, 137Cs is a harmful radioactive element that emits high-energy β particles and γ rays with a half-life of 30.2 years. 137Cs is difficult to extract because it is fixed to soil particles. For the volume reduction technology development of contaminated soil, this study tried to evaluate the irreversible Cs adsorption capacity of granite-originated soil. The soil sample used in the study was collected from C horizon of the soil developed in Mesozoic mica granite. The soil texture, mineralogy, organic content, pH, EC, cation exchange capacity (CEC), water-soluble cation and anion content of the soil samples were determined. A kinetic adsorption experiment and an isotherm adsorption experiment were performed to understand the overall Cs adsorption characteristics using 133Cs. The desorption of Cs by 0.1 mM KCl was also tested for the sample spiked with 133Cs and 137Cs. The soil sample showed a pH of 6.73, EC of 24.50 μS cm-1, and CEC of 1.34 cmolc kg-1, organic matter of 0.53% and sandy loam in texture. Quartz, feldspar and mica were identified as the major mineral components of bulk sample. The clay fraction consists of mica, hydroxyl-interlayer vermiculite (HIV), vermiculite and kaolinite. In the kinetic adsorption experiment, the Cs adsorption showed fast adsorption rates at the initial stage (6 hours) regardless of the 133Cs concentration, and the adsorption equilibrium state was reached after 48 hours. It was the most suitable for the pseudo second-order model. The 133Cs adsorption increased nonlinearly from low to high concentration, which was well match with the dual site Langmuir model. As a result of the desorption experiment, desorption was not performed up to 1.1 mg kg-1 in the presence of competitive ions K+, which is about 0.035% of CEC calculated by the isotherm model. The adsorption of Cs was controlled by frayed edge sites (FES) at a low concentrations and by basal sites or interlayer sites at a high concentration. Irreversible Cs fixation of by FES may be contributed by mainly weathered mica, and when these minerals are separated from the granite origin soil, the possibility of reducing the contamination concentration and volume of radioactive soil waste can be expected.

The integrity of the disposal repository structure must be guaranteed for few hundreds to few hundred thousand years until toxicity of radioactive waste is surely degraded. Acoustic emission (AE) method is widely utilized to evaluate the integrity of the structure because it can detect crack wave signals of the structures. It is well known that the cracking AE energy is proportional to the volume of the structure (Fractal theory). However, it is hard to destroy whole structures for obtaining AE energy. Therefore, the scaled specimens are prepared to obtain the relationship between volume of the structure and AE energy. The specimens are prepared with same of Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC) silo concrete recipe. Their diameters are from 50 mm to 150 mm in each 10 mm and their heights are twice of the diameter. One set of 50 mm to 150 mm specimens (11 specimens in one set) are made in single mixers to maintain uniformity. Surface of the specimens are flatten with cement milk to prevent from applying load with eccentricity. The uniaxial compression test is performed by controlling displacement as 0.1 mm/min. The fractal constant is obtained using least square function from volume-cumulative AE energy relationship.

Especially for near-surface repository for disposal of the low- and intermediate-level radioactive waste, safety assessment in case of inadvertent human intrusion should be handled seriously. This is because this type of incident will possibly give rise to high acute, not chronic exposure dose even though its occurrence of likelihood could higher than rather deeper geological repository for disposal of high-level radioactive waste over long time span after closure of the repository. Recently well drilling scenario for the pumping groundwater from the aquifer near the repository, among other possible inadvertent human intrusion incidents, has been popularly evaluated for the worst case due to its relatively high possibility of occurrence in parallel with normal scenarios for the nuclide transport for post-closure safety assessment of the repository. Movement of nuclide plume both in the confined and unconfined aquifer under and over a radioactive waste repository is of importance especially around an extracting well. Through this study a simple comment regarding quantification between a pumping rate from the well drilled into the aquifer as well as quantification of the plume size flowing around the well is presented. Drawdown of the well which is the change of water level of the upper water surface of the aquifer due to well pumping makes a cone of depression. And capture zone in the aquifer which is formed around the well, by which the groundwater is removed out, is the groundwater volume or area in the aquifer that is considered to contribute the extraction of the well by pumping. Usually this capture zone does not encompass the entire aquifer thickness for the partially penetrating well, which means that not all the portion of flowing groundwater through the aquifer is drawn by the well. And this capture zone does not need to coincide with the volume of the cone. Furthermore, all the nuclide plume volume is not necessarily and completely mixed with the groundwater flowing the entire aquifer. Therefore, a strategical approach might be required to grasp the aquifer portion and the plume size influenced by pumping to evaluate rather accurate radiological consequences due to the well scenario avoiding overestimation and meaningless conservatism as well, which is especially very common in the mass balance modeling e.g., by GoldSim under assumption that all the groundwater volume from the aquifer near the well extracted by the well. Although the capture zone around the well should be determined both by use of global/local groundwater flow model in the aquifer but a simple analytical model could be sought. Capture zone analysis has been widely seen in the area of the design of groundwater remediation system. If for safety assessment of the subsurface repository the plume behavior in the aquifer under the repository should be well characterized and correctly modeled, then the current study is expected to be more or less helpful to develop a specific mass balance model for nuclide transport and groundwater flow for assessment of an abnormal well drilling scenario near the repository.

Technetium (Tc) is a long-lived radioactive isotope, which exists as TcO4 - with high solubility under oxidative condition. The solubility of Tc is fundamental to assess the safety of radioactive waste repository in the case of a leakage of radioactive wastes. Cellulosic materials (paper, wood, cotton, etc.) contaminated by radionuclides are disposed of in low-level and intermediate-level radioactive waste repositories. Cellulose can be decomposed under anaerobic and alkaline conditions when cement pore water is saturated, and then isosaccharinic acid (ISA) is generated as a degradation product of cellulose. ISA forms complexations with radionuclides in solution and affects the solubilities of radionuclides. Therefore, the effect of ISA should be accurately evaluated to predict and assess the mobility of radionuclides in repository environments. In this study, batch tests were conducted to confirm the effect of ISA on the solubility of Rhenium(IV) Oxide. Herein, rhenium was used as a non-radioactive analog of Tc due to their similar chemical properties. Deionized water (DIW) and 0.1 M NaOH solution in pH 12.5 were used as background solutions, and ISA concentration was varied to 1~20 mM using Ca(ISA)2 and NaISA, respectively. The batch tests were conducted under both aerobic and anaerobic conditions. The whole batch tests under anaerobic conditions were performed in the glove box using oxygen purged DIW with a high purity nitrogen gas (99.9%) and low oxygen concentration (< 0.5 ppm). As a result, the rhenium concentration decreases as more ISA is dissolved in the solution, which shows the contrary effect of ISA on the solubility of other metals and radionuclides (e.g., Co, Th, Fe, Ni, etc.). It is assumed that the reducing capacity of ISA decreases the rhenium dissolution in the solution. Additional characterization of the oxidation state of rhenium oxide and the mechanism will be tested and presented.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.