Molten salt is one of the promising medium materials for molten salt reactors and energy storage systems. Molten salt is advantageous for better physical properties such as low melting point and high boiling point, high energy capacity, high thermal conductivity, and high thermal stability than other medium materials such as water or liquid metals. However, the corrosivity of the molten salt is one of the main factors that disturbs the various applications of the molten salt. On the other hand, metallic 3-D printing technologies have developed by leaps and bounds over the past 20 years and show potential for use in cutting-edge industries such as aerospace and military purposes. However, the biggest problem of 3-D printed products is that the mechanical and physical properties are very weak along the laminated plane that was generated during the manufacturing process. In particular, other research showed that corrosion is vulnerable through the laminated surface, and corrosion along the laminated plane is not completely mitigated through a general heat treatment process although the microstructure of the surface is evaluated to be partially mitigated by the heat treatment. In this study, molten salt corrosion behaviors of simple Ni-based alloy with a composition of 80Ni- 20Cr were analyzed. Ni-based alloys were fabricated by casting and 3-D printing, and some of the 3-D printed specimens were thermally treated at 1,273 K for 1 hour to examine the effects of heat treatment on corrosion behaviors. In molten eutectic NaCl-MgCl2 melts at 973 K, Ni-based alloys were corroded for 1, 3, 7, and 28 days and their microstructural changes were analyzed by SEM-EBSD-EDS and OM. The corrosion behaviors of the alloy were also evaluated by the salt composition measured with ICPOES. 3-D printed alloy with post-treatment showed more resistivity to the molten salt corrosion than as-fabricated 3-D printed alloy. However, the corrosion rate of the 3-D printed specimen after heat treatment was still higher than that made by casting.

Zirconium(Zr) alloys are commonly used in the nuclear industry for applications such as fuel cladding and pressure tubes. To minimize the levels and volumes of radioactive waste, molten salts have been employed for decontaminating Zr alloys. Recently, a two-step Zr metal recovery process, combining electrolysis and thermal decomposition, has been proposed. In the electrolysis process, potentiostatic electrorefining is utilized to control the chemical form of electrodeposits(ZrCl). Although Zr metals are expected to dissolve into molten salts, reductive alloy elements can also be co-dissolved and deposited on the cathode. Therefore, a better understanding of the anodic side’s response during potentiostatic electrorefining is necessary to ensure the purity of recovered Zr and long-term process operation. As the first step, potentiodynamic polarization curves were obtained using Zr, Nb, and Zr-Nb alloy to investigate the anodic dissolution behavior in the molten salts. Nb, which has a redox potential close to Zr, and Zr exhibit active or passivation dissolution mechanisms depending on the potential range. It was confirmed that Zr-Nb alloy also has a passivation region between -0.223 to -0.092 V influenced by the major elements Zr and Nb. Secondly, active dissolution of Zr-Nb was performed in the range of -0.9 to -0.6 V. The dissolution mechanism can be explained by percolation theory, which is consistent with the observed microstructure of the alloy. Thirdly, passivation dissolution of Zr, Nb, and Zr-Nb alloy was investigated to identify the pure passivation products and additional products in the Zr-Nb alloy case. K2ZrCl6 and K3NbCl6 were identified as the pure passivation products of the major elements. In the Zr-Nb alloy case, additional products, such as Nb and NbZr, produced by the redox reaction of nanoparticles in the high viscous salt layer near the anode, were also confirmed. The anodic dissolution mechanism of Zr-Nb alloy can be summarized as follows. During active dissolution, only Zr metal dissolves into molten salts by percolation. Above the solubility near the anode, passivation products begin to form. The anode potential increases due to the disturbance of passivation products on ion flow, leading to co-dissolution of Nb. When the concentration of Nb ion exceeds the solubility, a passivation product of Nb also forms. In this scenario, a high viscous salt layer is formed, which traps nanoparticles of Zr metal, resulting in redox behavior between Zr metal and Nb ion. Some nanoparticles of Zr and Nb metal are also present in the form of NbZr.

Korea Atomic Energy Research Institute (KAERI) is planning to disposal of the radioactive contaminated cement waste form to the final disposal facility. The final disposal facility require evaluation of immersion, compressive strength, and radionuclide inventory of radioactive wastes to meet the acceptance criteria for safe disposal. According to the LILW acceptance criteria of the Nuclear Safety and Security Commission ok Korea (NSSC), the disposal limit radioactivity of 129I (3.70×101 Bq/g) is lower than other radionuclides. 129I emits low energy as its disposal limit is low, so it is difficult to analyze in the presence of 137Cs and 60Co which emit high energy. Therefore, it is essential to an accurately separate and analyze iodine in radioactive waste. In this study, we focused on the determination of 129I in cement waste form containing 137Cs, 60Co. We added 1 g of 129I(11.084 Bg), 137Cs(1,300 Bq) and 60Co(402 Bq) to cement waste form, respectively. The separation of 129I in cement waste form was carried out using an acid leaching method. And, we confirmed the specific activity of 137Cs and 60Co at each separation step. It was observed that an acid leaching method showed the remove efficiency 137Cs(99.97%) and 60Co(99.94%), respectively. In addition, 129I was also analyzed at approximately 96.44% in simulated contaminated cement waste form. In conclusion, through this experiment, it was confirmed that 129I could be successfully separated and analyzed by using the acid leaching method in cement waste form containing excessive 137Cs and 60Co.

A molten salt reactor (MSR) is a conceptual nuclear reactor that uses molten salt with liquid fuel as its primary coolant. Based on the thermophysical and neutronic properties, MSR has advantages such as high efficiency, safety, combustion of transuranic (TRU) elements, and availability of miniaturization and on-power refueling. Various research on MSR such as system development, neutronic analysis, material development, and molten salt property analysis has been conducted, but the biggest problem is the molten salt corrosion. The molten salt corrosion on structural materials can be explained by two processes; electrochemical and chemical reactions. The reduction of oxidative ions such as fuel and TRU elements is one of the major causes of molten salt corrosion. Contamination by humidity and oxygen is also known as the accelerating factor of molten salt corrosion. Also, molten salt corrosion behaviors on structural material deteriorate when dissimilar alloys are introduced in the molten salt system. Various techniques to mitigate molten salt corrosion in fluoride system has been developed, but these are not well-verified in chloride system. In this research, various methodologies to mitigate molten salt corrosion are studied. The corrosion behaviors of 80Ni-20Cr alloy in molten eutectic NaCl-MgCl2 salt at 973 K are analyzed with various applications such as salt purification, sacrificial metal injection, and salt redox potential control. Oxygen and water impurities that can accelerate molten salt corrosion have been removed by electrochemical and chemical methods; Applying the reduction potential for H+/H2 and oxidation potential for O2-/O2, introducing HCl and CCl4 gas, and introducing the metallic Cr and recovering the ionized Cr. Corrosion acceleration/deceleration effects were analyzed when introducing the reducing reagent such as Mg and Nb or oxidizing reagent such as metallic Mo and the effect of inert metallic element (W) was also investigated. The salt potential was controlled by applying the potential to the salt and adjusting the Eu3+/Eu2+ ratio.

Low- and intermediate-level radioactive waste for permanent disposal often contains organic complexing agents, so-called chelating agents. Organic complexing agents, which are polycarboxylic acids, can increase the mobility of radionuclides into the environment by forming water-soluble complexes with most heavy metals. Therefore, analyzing the complexing agents in radioactive waste is crucial for comprehensive management of nuclear wastes. According to regulatory guidelines, specifically Notice No. 2021-16 issued by the Nuclear Safety and Security Commission, the determination of chelating agent content in radioactive waste materials is required to ensure proper management and safe disposal. However, only a few methods are available to analyze the chelators in various matrices such as concrete, metals, soil, and mixed solid wastes like plastics, vinyl, and rubber. Recently, we found a UV-Vis method based on an enzymatic reaction is inadequate for analyzing citric acid in radioactive waste with a complex matrix like concrete. To address this, we developed a method to determine the contents of EDTA and NTA using a UV-Vis spectrophotometer and citric acid using ion chromatography. The results showed good validity and reliability to determine the chelating agents in various radioactive wastes.

For decontamination and quantification of trace amount of tritium in water, an efficient separation technology capable of enriching tritium in water is required. Electrolysis is a key technology for tritium enichment as it has a high H/T and D/T separation factors. To separate tritium, it is important to develop a proton exchange membrane (PEM) electrolyzer having high hydrogen isotope separation factor as well as high electrolyzer cell efficiency. However, there has not been sufficient research on the separation factor and cell efficiency according to the composition and manufacturing method of the membrane electrode assembly (MEA) Therefore, it is necessary to study the optimal composition and manufacturing method of the MEA in PEM electrolyzer. In this study, the H/D separation factor and water electrolysis cell efficiency of PEM electrolyzer were analyzed by changing the anode and cathode materials and electrode deposition method of the MEA. After the water electrolysis experiment using deionized water, the D/H ratio in water and hydrogen gas was measured using a cavity ring down spectrometer and a mass spectrometer, respectively, and the separation factor was calculated. To calculate the cell efficiency of water electrolysis, a polarization curves were obtained by measuring the voltage changes while increasing the current density. As a result of the study, the water electrolyzer cell efficiency of the MEA fabricated with different anode/cathode configurations and electrode formation methods was higher than that of commercial MEA. On the other hand, the difference in H/D separation factor was not significant depending on the MEA fabrication methods. Therefore, using a cell with high cell efficiency when the separation factor is the same will help construct a more efficient water electrolysis system by lowering the voltage required for water electrolysis.

The large rectangular and cylindrical concrete drums are stored in nuclear power plant (NPP) for a long time. At the early stage of NPP operation, the treatment technology of boron concentrates and spent resin was not well developed, when compared to current system. Since the waste acceptance criteria (WAC) of the disposal facility was not established, the boron concentrates and spent resins were packaged in 200 L drum. Some of the 200 L drums, which contain relatively high dose rate radioactive waste, were stored in large concrete drum. The concrete drum offers superior shielding effect and allows reduction of radiation exposure to workers. The WAC requires various characteristics: radiological characteristics, physical characteristics, chemical characteristics, etc. The non-destructive method allows the rapid evaluation and estimation of the concrete structure. Also, it is expected that the large concrete exhibits integrity after the measurements. In this paper, the non-destructive method to understand the large rectangular and cylindrical drum is systematically studied. The advantage and disadvantage of the non-destructive methods were compared in this paper. In addition, the optimized methodology to characterize the radioactive waste containing large rectangular and cylindrical drum will be discussed in this paper.

During the decommissioning of a nuclear power plant, the structures must be dismantled to a disposal size. Thermal cutting methods are used to reduce metal structures to a disposal size. When metal is cut using thermal cutting methods, aerosols of 1 μm or less are generated. To protect workers from aerosols in the work environment during cutting, it is necessary to understand the characteristics of the aerosols generated during the cutting process. In this study, changes in aerosol characteristics in the working environment were observed during metal thermal cutting. The cutting was done using the plasma arc cutting method. To simulate the aerosols generated during metal cutting in the decommissioning of a nuclear power plant, a non-radioactive stainless steel plate with a thickness of 20 mm was cut. The cutting condition was set to plasma current: 80 A cutting speed: 100 mm/min. The aerosols generated during cutting were measured using a highresolution aerosol measurement device called HR-ELPI+ (Dekati®). The HR-ELPI+ is an instrument that can measure the range of aerodynamic diameter from 0.006 μm to 10 μm divided into 500 channels. Using the HR-ELPI+, the number concentration of aerosols generated during the cutting process was measured in real-time. We measured the aerosols generated during cutting at regular intervals from the beginning of cutting. The analyzed aerosol concentration increased almost 10 times, from 5.22×106 [1/cm3] at the start of cutting to 6.03×107 [1/cm3] at the end. To investigate the characteristics of the distribution, we calculated the Count Median Aerodynamic Diameter (CMAD), which showed that the overall diameter of the aerosol increased from 0.0848 μm at the start of cutting to 0.1247 μm at the end of the cutting. The calculation results were compared with the concentration by diameter over time. During the cutting process, particles with a diameter of 0.06 μm or smaller were continuously measured. In comparison, particles with a diameter of 0.2 μm or larger were found to increase in concentration after a certain time following the start of cutting. In addition, when the aerosol was measured after the cutting process had ended, particles with a diameter of 0.06 μm or less, which were measured during cutting, were hardly detected. These results show that the nucleation-sized aerosols are generated during the cutting process, which can explain the measurement of small particles at the beginning of cutting. In addition, it can be speculated that the generated aerosols undergo a process of growth by contact with the atmosphere. This study presents the results of real-time aerosol analysis during the plasma arc cutting of stainless steel. This study shows the generation of nucleation-sized particles at the beginning of the cutting process and the subsequent increase in the aerosol particle size over time at the worksite. The analysis results can characterize the size of aerosol particles that workers may inhale during the dismantling of nuclear power plants.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

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

Several tests should be performed to estimate the structural and chemical stability of the radioactive waste. Among the tests in Gyeongju LILW repository, the leaching test which follows ANS 16.1 standard test method should be conducted for Cs, Sr, and Co radionuclides and must satisfy leachability index larger than 6 which applies deionized water as a leachant. However, the expected leachant inside the silo is groundwater that contains various ions and a high pH condition is predicted due to the concrete structures inside the silo. According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. Cobalt precipitates when the leachant has high pH environment which is similar condition to the cement-saturated leachant. Unlike the cobalt, cesium is preferred to exist as ion in the high pH condition. Therefore, the significant effect of the chemical environment of the leachant on the leachability of the radionuclides should be considered for the waste acceptance criteria of the radioactive waste repository. From the ‘NRC, Technical position on the waste form, rev1’, the leaching test method should follow the ANS 16.1 methods by using deionized water as leachant, however, a new leachant showing more aggressive leachability can be applied instead of deionized water. In the other hand, ASTM C1308 leaching test method recommends applying actual groundwater of the repository as a leachant. FT-04-020, the leaching test method of France, suggests the ion composition of the groundwater including the pH value. Therefore, the adequacy of using deionized water as leachant for the leaching test method of Cs, Sr, and Co should be re-examined. In this study, the leaching behavior of Cs, Sr, and Co under the several leachant types is estimated. The cement solidified specimen containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water, simulated groundater, and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type was discussed. The result of this study is expected to be a background data that helps understanding the actual leaching behavior of the Cs, Sr, and Co in the Gyeongju LILW repository.

To prevent the release of radionuclides into the biosphere, disposal facilities for radioactive waste should be located to provide isolation from the accessible biosphere for tens of thousands to a million years after closure. During the period of interest, the constantly evolving natural environment and possible geological events of the site can cause disturbances to the containment function of the repository. Thus, for the long-term safety assessment of the repository, the possible long-term change of natural barrier should be considered. Due to the characteristics of radionuclides that transport mainly through the groundwater, understanding the long-term evolution of groundwater flow and geochemical properties is essential to assess the long-term changes in the natural barrier performance. The changes in characteristics of natural rocks and geological structures are one of the main factors that determine the hydrological and geochemical characteristics of the deep underground. In this study, we plan to develop a methodology to estimate these future geological evolutions in order to assess the possibility of hazardous events of the site that can affect hydrological or geochemical properties over the period of interest, and also in order to verify the change in the geological environment is within the safe performance range even after the period of interest. However, it is very unreliable to predict future changes in the natural environment because it is very heterogeneous, complex, and difficult to observe directly. For the preliminary study of the project, we reviewed cases of future evolution prediction researches with regard to the geological environment of disposal site and methods they applied to reduce the uncertainty of the prediction. The results will be used to establish basic data for future studies on the long-term evolution of hydraulic-mechanics performance of natural barrier and long-term evolution of geochemical performance around KURT site. In addition, it can contribute to construct long-term evolution scenario of the geological environment around future URL site.

Long-term evolution of the surface environments can affect the safety of deep geological disposal. Therefore, it is important to understand the water balance components constituting the water cycle among atmosphere, surface, and subsurface. In Finand, the surface and near-surface hydrological model (SHYD) was developed to calculate the water balance of Olkiluoto Island. Through the intensive site investigations, the data sets as input for the site scale model in present-day conditions have been collected such as transpiration and meteorological data. In this study, weighing lysimeter method was selected to quantify small-scale soil water balance of the vadose zone in the UNsaturated zone In-situ Test facility (UNIT) around KAERI Underground Research Tunnel. Hydrological components such as precipitation, evapotranspiration (ET) and leachate were derived from water balance analysis on the lysimeter measurements in UNIT. Among the hydrological components, actual ET accounts for more than 50% of the annual precipitaion, and thus plays an important role on predicting the hydrological evolution in the future. In this context, actual ET measured from the weighing lysimeter was compared with potential ET estimated from meteorological data using FAO-56 Penman-Monteith method.

A radioactive waste repository consists of engineered barriers and natural barriers and must be safely managed after isolation. Geologic events in natural barriers should be categorized and evaluated according to their magnitude to assess the present and future stability of disposal. Among the longterm evolutionary elements of natural barriers, faults are a small portion of the Earth’s crust. Still, they play an important role in nuclide transport as conduits for fluids moving deep underground. In addition, the physical and chemical properties of fault rocks are useful for understanding the longterm and short-term behavior of faults. Paleomagnetic research has been used extensively and successfully for igneous, metamorphic, and sedimentary rocks. In addition, magnetic characterization of fault rocks can be used to describe faults or infer the timing of major geological events along fault zones. Components of magnetization defined in fault-breccias were attributed to chemical processes associated with hydrothermal mineralization that accompanied or post-dated tectonic activity along the fault. The study of magnetic minerals in fault rocks can be used as “strain indicators”, “geothermometers”, etc. This study is a preliminary test of magnetic properties using fault gouges. Fault gouges are not well preserved in typical terrestrial environments. Access to fresh gouges typically requires trenching through faults or sampling with a core drill. Fortunately, it is a magnetic property study using a fault gouge that exists on the inner wall of KURT (KAERI Underground Research Tunnel). This is to identify the motion history of the fault and, furthermore, to understand the stress structure at the time of fault creation. In addition, it can be presented as evidence for evaluating faults that may appear in future URL (Underground Research Laboratory).

Surface environmental factors such as climate change can affect the safety of the disposal system by changing groundwater recharge or flow. Therefore, it is important to identify surface environmental factors and hydrogeological factors to evaluate long-term changes in hydrogeological environment of a disposal system. In particular, evapotranspiration is an important to be considered because it loses 70% of rainfall and has a great effect on groundwater recharge. Evapotranspiration can be estimated using simple or complex models based on meteorological data. Meteorological data from January 2010 to December 2022 were collected from 44 Automatic Synoptic Observation Systems (ASOS) of the Korea Meteorological Administration (KMA), which observe factors necessary for calculating evapotranspiration. For the estimation of evapotranspiration through simple models, temperature-based models (Blaney-Criddle method, modified Blaney-Criddle method, Hargreaves-Samani method) and radiation-based models (Simple Abtew method, Makkink method, Prietley-Taylor method, Turc method, Solar radiation-Maximum temperature method) were used. The calculation of evapotranspiration through the complex model used the Penman-Monteith method, which is used as a standard model in the USA, Japan, and FAO. By comparing the evapotranspiration calculated by complex and simple model, methods with small errors were identified each region. In addition, long-term climate change scenarios were applied to confirm changes in long-term evapotranspiration in South Korea. The results of this study will be used to find alternative models in the case of missing data in the Penman-Monteith model, which requires a lot of meteorological data, and can be used as basic data for calculating groundwater recharge that can affect the disposal system in the future.

Performance and safety assessments for deep geological disposal are often conducted over a longterm time scale, such as from hundreds of thousands to a million years. During this period, it is expected that the surface environment will be changed significantly. Uplift-subsidence and erosion-deposition are thought to be included as the main causes of the changes, and it is necessary to evaluate their expected effects. In this study, the conceptual processes of the changes in the surface environment components were to be presented by identifying the uplift-subsidence and erosion-deposition processes and analyzing their effect on the surface environment components. For inferring the long-term change process of the surface environment due to the internal activities of the Earth, the process of uplift and subsidence caused by crustal movements that change the subsurface environment through the deep and sallow underground was briefly presented in the form of a chain flowchart. Uplift-subsidence is mainly caused by diastrophism due to tectonic movement, such as subduction at the boundary of plates. They can change the geomorphology by affecting sealevel change and erosion-deposition. The changed geographical features have an influence on the distribution of surface water and the flow path of groundwater. They also have an impact on the scale and processes of local uplift and erosion, which can be the main factors of pedogenesis and vegetation in the local site. The results of this study can be helpful for formulating scenarios related to long-term evolution in the surface environment required for performance and safety assessments of deep geological disposal.

In KAERI, a site descriptive model for stress field estimation had already been constructed by using integrated field data within KURT site scale. A sub-divided rock block domain containing major fracture zones has spatial rock mass and fault properties. The properties were decided based on the rock classification results of several borehole investigations. Modeling for maximum and minimum horizontal stress field estimation was performed and compared with the in-situ data. As a result, a depth-dependent stress ratio was adopted to obtain numerical results closer to actual in-situ data. Although the results were suitable at a relatively low depth (~500 m), there is still some deviation trend at a deep depth. This study aims to improve these modeling results by incorporating not only depth-dependent stress ratio but also changes in rock mass properties along the depth. The deep borehole of DB2 in the KURT site indicated fracture distribution corresponding to the property changes. Natural fractures are typically randomly oriented, and the fracture frequency decreases with increasing depth. The increase in P-wave velocity log data accompanies these features. A discrete fracture network (DFN) model can be used to simulate fractured rock explicitly, but DFN modeling is not feasible for site scale analysis because of its numerical efficiency. Therefore, as a preliminary model in this study, the effect of fracture distribution was considered by substituting the influence for the depth-dependent property. The properties were estimated from the fracture frequency and P-wave velocity log data. The influence of elastic modulus and density on the site stress field was dominant, with decreasing the deviation trend between modeling and in-situ data at a deep depth. Considering that the depth of the repository construction is within about 500 m, it may not be necessary to consider the change of rock properties with depth. However, it was determined that the rock property effect might need to be considered when the loading conditions change due to subsidence in the long-term evolution scenario. Continuously, this site descriptive modeling will be interdependently conducted with a representative DFN block model for deriving equivalent properties in fractured rock.

Two sets of analyses for the cases of groundwater release to well and sea ecosystems were conducted for the environmental impact assessment of high-level radioactive waste disposal facilities. After obtaining the respective BDCF (Biosphere Dose Conversion Factor) results for the scenarios of well-farming and marine water fishing using different biosphere assessment conceptual models implemented in ECOLEGO, they were compared each other. The purposes of these analyses are to identify reference generic biosphere conceptual models and to get insight on model uncertainty. In this study, the endpoint used for the comparison of the ECOLEGO biosphere models was the socalled Biosphere Dose Conversion Factor (BDCF), which is defined as the maximum value of the total dose to the exposed group, in Sv/yr, resulting from a continuous unit release of 1 Bq/yr during the whole simulation time either to the well compartment (BDCF_Well) or to the marine water compartment (BDCF_Sea). The radionuclides considered in the comparison were Cs-137, I-129, Nb-94, Ni-59, Ni- 63, Sr-90 and Tc-99. The conceptual models used in the biosphere assessment of the releases to a well are based on models that have been used by the DOE (simple-soil model) and SKB (complex-soil model) in safety assessments of radioactive waste repositories, respectively. Difference between two conceptual models used in the assessment of the releases to a sea is the number of compartments representing the sea; i.e., one model represents the sea with one compartment for the water and one for the sediment (singlecompartment model), whereas the alternative model uses two compartments for the water and the sediments: one for the inner coast and one for the outer coast (double-compartment model). The results of the BDCF_Well to a farmer obtained with the DOE and SKB models are shown to be very close to each other. Despite the differences in conceptual models and parameters, the results are within a maximum difference of a factor of 4. The results from the SKB model were higher for all radionuclides. The values of the BDCF_Sea obtained with the single- and double-compartment models are shown to be larger differences with a maximum order of 2. For all studied radionuclides, the double-compartment model produces higher BDCFs than does the single-compartment model. The differences would be due to activity concentrations in both water and sediments. Since the hydrodynamic behavior assumed for flow in the sea could significantly influence the dilution volumes and hence the concentrations, it is found that site-specific investigations are necessary to establish an appropriate marine biosphere conceptual model.

The distribution characteristics of rock fractures determine the hydro-mechanical behavior of natural barriers. Rock fractures are defined by various parameters, which are analyzed as the probability distribution from observation results by surveying the exposed rock surface or borehole. The size is known to have the most uncertainty among the fracture parameters because it cannot be directly measured. Therefore, various estimation methods have been proposed for fracture size distribution using the fracture traces observable on the rock surface. However, most methods are based on a planar survey area, limiting their applicability to the underground research laboratory (URL) excavated in the form of tunnels. This study aims to review a method that can be applied to estimate the size distribution of fractures in deep rock masses at the URL site. The estimation method using the joint center volume (JCV) has recently been extended to be applicable regardless of the geometry of the survey area, which means that it can be applied to the URL site with complex structures. To apply the JCV-based estimation method to non-planar survey areas, JCV calculation using Monte Carlo simulation and estimation of fracture size distribution using the maximum likelihood method are required. In this study, we applied the JCV-based estimation method to a tunnel-shaped survey area to examine its applicability to the URL site. The error rates were analyzed when there were fracture sets with various orientations, size distributions, and maximum fracture sizes in the rock mass, and it was found to be less than 10% in all cases. This result indicates that the JCV-based estimation method can be used to estimate the fracture size distribution of the surrounding rock mass if accompanied by a reliable survey of fracture traces on the tunnel surface inside the URL site. Also, since there are no restrictions on the geometry of the survey area, we can continuously update the estimation results during the URL excavation process to increase reliability. The fracture size distribution is essential for constructing the discrete fracture network (DFN) model of the rock mass units at the URL site. In the future, the uncertainty for the fracture size in the DFN model is expected to be reduced by applying the JCV-based estimation method.

The deep geologic repository (DGR) concept is widely accepted as the most feasible option for the final disposal of spent nuclear fuels. In this concept, a series of engineered and natural barrier systems are combined to safely store spent nuclear fuel and to isolate it from the biosphere for a practically indefinite period of time. Due to the extremely long lifetime of the DGR, the performance of the DGR replies especially on the natural geologic barriers. Assessing the safety of the DGR is thus required to evaluate the impacts of a wide range of geological, hydrogeological, and physicochemical processes including rare geological events as well as present water cycles and deep groundwater flow systems. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the numerical models used for safety evaluation need to be comprehensive, robust, and efficient. This study describes the development of an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be approved with confidence by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM under development can currently simulate overland flow, groundwater flow, near surface evapotranspiration in a modular manner. The IHM can also be considered as a framework as it can easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using illustrative examples.