In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

The background of the development is to contribute to the reduction of radioactive waste, recycling of resources and effectively purifying the air in the workplace. Ultimately, it affects the reduction of internal exposure of workers. According to the standard procedure of KHNP,「Use and Management of Respiratory Protection Equipment」, the expiration date of mask filter is indicated by the manufacturer before opening. It is 1 year from the date of first combination after opening. We have developed an air purifying equipment that can recycle and reuse expired mask filter waste in nuclear power plant. In order to confirm the performance, we observed air pollution level by operation time. The location was measured at 3 locations including the decontamination product warehouse in NPP, and the size of the measurement locations were 53 m3, 150 m3, 180 m3, and 900 m3. As a result of measurement, significant air purification effect was found in 53 m3 and 150 m3. Decontamination effect of 80% was shown after 1 hour of operation, and 20% of decontamination effect was shown gently for 3 hours thereafter. On the other hand, there was no significant decontamination effect in the 180 m3 and 900 m3 spaces. Significant results were derived by statistical methods. Statistical procedure involves the collection of data leading to test of the relationship between two statistical data sets, or a data set and synthetic data drawn from an idealized model. The basic composition and product characteristics was as follows: Blower, filter fixing unit, Air purifier outlet round shape, Differential pressure gauge, inverter (200 V, 3π, 200 W). The developed product weigh is 25 kg. This is lighter than the existing product weighing 100 kg. It is judged that it is suitable for convenient use. Because the area where the major air pollution level occurs is isolated by a room in NPP. This developed product has a greater significance in that it recycles radioactive waste within the radiation management area rather than air purification efficiency.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

Low-and intermediate level waste (LILW) should be solidified and satisfy the waste acceptance criteria (WAC) to be disposed of in the LILW repository. The LILW should be uniformly solidified and should maintain its structural stability under the expected condition according to the WAC. Compressive strength of cement solidified waste should satisfy at least 3.44 MPa to be disposed of in the repository. In addition, its compressive strength should satisfy at least 3.44 MPa after the irradiation, immersion and leaching test. The compressive strength test and dimension of test specimen differ according to countries. However, measured compressive strength of solidified waste is affected by geometry of specimen and test condition. Diameter, ratio between diameter and height, and porosity are one of factors that affect to the compressive strength of cement solidified waste. Generally, specimen with larger diameter shows higher value of measured compressive strength. The ratio of height and diameter shows similar tendency to the diameter while larger porosity generally lowers the compressive strength. In other hands, higher compressive strength is expected when the loading rate is higher during the compressive strength test. U.S. is applying loading rate from ASTM C39 (0.25±0.05 MPa) for the compressive strength test while Korea is applying loading rate from KS F 2405 (0.6 MPa·s−1). France applies loading rate following FT-02-010 (0.5 MPa·s−1) for cement solidified waste. As the measured compressive strength increases when the loading rate increases, the effect of loading rate to the compressive strength of cement solidified waste should be assessed by quantification and consider its effect on the sight of regulation. In this study, the effect of geometric parameters of specimen and test condition to the compressive strength are checked by manufacturing specimen by solidifying mock sludge waste with cement. To prevent increasing amount of secondary waste, effects of ratio of height and diameter and porosity to the compressive strength are checked while diameter value is fixed. For loading rate, loading rate from ASTM C39 and KS F 2405 were compared. Existence of significant variance of measured compressive strengths of cement solidified waste are check by performing statistical analysis. Finally, by analyzing the relationship between test condition and measured compressive strength, the test method that measures the compressive strength conservatively is aimed to be derived.

During the treatment of spent nuclear fuel, radioactive iodine is generated in a liquefied or gaseous form in a specific process. In the case of iodine 129, it is a long-lived nuclide with a very long halflife and has high groundwater mobility under repository conditions. Despite showing a low radioactivity value, research on the management of radioactive iodine from a long-term perspective is continuously being performed. Although research has been conducted using borosilicate glass as a medium for solidifying iodine, compatibility of I in borosilicate glass is very small and the volatility is high in the solidification process. So it is not suitable as a solidified substance of iodine. Therefore, studies on other solidification media to replace them are continuously being conducted. Our research team tried to develop a new medium that can contain iodine in a solidified body stably through a simple heat treatment process and can improve problems such as volatility and waste loading. Iodine is captured as AgI in the Ag ion-exchanged zeolite. So, TeO2, Ag2O, and Bi2O3 having a high AgI loading rate were used as main components. It was named TAB after taking the first letter of each element. In previous studies, the physical properties, structure, and chemical stability of TAB materials were confirmed. PCT (Product Consistent test) was performed to confirm chemical stability. It is mainly used to compare the chemical stability of glass materials with other glass materials, but there are limitations in evaluating the long-term chemical stability of materials. In this experiment, we tried to evaluate the long-term stability of TAB and compare it with borosilicate, which is conventionally used to treat radioactive waste. In addition, we tried to understand the leaching behavior inside the TAB medium. For this purpose, ASTM C1308 test was performed for 365 days, and distilled water and KURT groundwater were used as leachates to examine the effect of ions in the groundwater on the solidified body. To analyze the leaching behavior, ICP-MS and ICP-OES analyses were performed, and the cross-section of the sample after leaching was observed through SEM.

In Korea, it is expected that the decommissioning of nuclear reactors will increase due to the license termination of reactors constructed in the 1960s to the 80s. According to the investigation of KORAD, VLLW accounts for 67.10% of decommissioning wastes and amounts to about 413,336 drums. Due to their huge amount, it is necessary to create an appropriate decommissioning waste management plan even though VLLW is disposed at the second-phase disposal facility of the Gyeongju repository. For efficient reduction in decommissioning wastes, it is required to actively use a clearance of metallic and concrete radioactive wastes. Regulations of nuclear safety and security commission notice that the radioactive waste can be reused or recycled if it meets the clearance criterion, 10 μSv·y−1 for individual dose. Therefore, it is important to develop a computational code which calculate individual doses for each scenario, and determine whether the clearance criterion is satisfied. However, in the case of metallic waste, RESRAD-RECYCLE used in dose assessment for the clearance has no longer been maintained or updated since 2005 and there is no code for recycling of concrete waste. For this reason, a dose assessment code RUCAS (Recycle-Underlying Computational dose Assessment System) has been developed by Ulsan National Institute of Science and Technology (UNIST). A point kernel method is adopted into external dose assessment model to calculate more realistic options, which are various geometries of source, and shielding effect. In the case of internal radiation, equations of internal dose from IAEA are used. This research conducts a verification of dose assessment model for recycling of metallic radioactive waste. RESRAD-RECYCLE is the comparison object and results from RESRAD-RECYCLE validation report are referenced. Targets are 14 recycling scenarios composed up to the smelting metal step of four steps, which are arising scrap metal, smelting scrap metal, and fabrication of metal product, and reusing/recycling of product. Seven isotopes, which are Ac-227, Am-241, Co-60, Cs-137, Pu-239, Sr- 90, and Zn-65, are selected for calculation. Validation results for external dose vary by isotopes, but show acceptable differences. It seems to be caused by difference in the calculation method. In the case of internal dose using same calculation formula, results are exactly matched to RESRAD-RECYCLE for all isotopes. Consequently, RUCAS can conduct functions supported by RESRAD-RECYCLE well and future work will be conducted related to domestic recycling scenarios considering public acceptance, and verification with radiation shielding codes for various geometries of source.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.

The natural barrier, a component of the deep disposal system, has site-specific characteristics depending on the site of the repository, and is one of the main considerations for long-term safety evaluation after closure along with the engineered barrier among the multiple barrier systems of the repository. The natural barrier is defined in Korea as the natural underground and surface structures that can restrict the exposure of radioactive waste, human intrusion or groundwater infiltration into a disposal facility, and the transfer of radionuclides. It includes bedrocks and soils surrounding the engineered barriers of radioactive wastes [Notice of the NSSC, No. 2020021]. This study analyzed foreign regulatory requirements related to natural barriers, requirements for natural barrier and performance target of Sweden and Finland (safety functions and target characteristics of natural barriers, e.g. natural barrier composition, geological characteristics, hydrogeological characteristics). Overseas regulations and cases referenced to derive regulations of general safety requirements on natural barrier are IAEA SSG-14, SSMFS 2008:21 in Sweden, STUK/Y/4/2018 in Finland, and POSIVA SKB Report 01, a joint report between POSIVA and SKB. The repository site and repository depth should be chosen so that the geological formation provides adequately stable and favorable conditions to ensure that the repository barriers perform as intended over a sufficient period of time. The conditions intended primarily concern temperature- related, hydrological, mechanical (for example, rock mechanics and seismology) and chemical (geochemistry, including groundwater chemistry) factors. Furthermore, the repository site should be located at a secure distance from natural resources exploited today or which may be exploited in the future [SSMFS 2008:21]. Finland regulations also suggests similar requirements [STUK Y-4-2018]. According to the above regulations, POSIVA SKB report 01 mentions both the host rock and the underground opening as natural barriers and requires a safety function, and the main safety functions of the host rock and underground opening are as follows: (1) Isolation from the surface environment; (2) Favorable thermal conditions; (3) Mechanically stable conditions; (4) Chemically favorable conditions; and (5) Favorable hydrogeological conditions with limited transport of solutes. Such safety functions would provide insight for understanding of the natural barrier of deep geological disposal system.

It can take hundreds of thousands of years for decreasing radiological effects of high-level radioactive wastes to those of natural background radiation. Therefore, long-term time scale should be considered in order to demonstrate performance and safety of deep geological disposal of the radioactive wastes. The changes of surface environment for these long-term time scale can have influence on safety analysis by changing transport path of radionuclides from the radioactive wastes. Changes in climate is considered as one of main factors causing the long-term changes of the surface environment. The own effects and interactions of climate with other components of the geological disposal system are organized in features, events, and processes (FEPs). In this study, some natural processes occurred by changes of climate were suggested and the connectivity between each process is proposed based on the relation of the FEPs concerned with the changes of climate and surface environment. The processes were classified into global and regional/local scales and was analyzed, respectively. Then, the influences of the processes on shallow groundwater and surface water body environment, which might be transport path of radioactive nuclides in local/site scales, were expected. As the proposed connection demonstrate the order or hierarchical relations of the natural processes, it can shows that some output by a certain process may be input of other process connected the former process in numerical simulations for interpreting the processes. If the connection may be considered to be suitable to represent longterm changes of the surface environment, it can be evaluated that the expected scenarios based on the connection is also proper. In addition, it can be helpful in selecting factors to be studied more detailed in terms of climate change for expecting long-term changes in the surface environment by analysis on the input and output data. The results of this study can be used as basic approaches to represent the long-term changes in the surface environment caused by specific natural processes from changes of climate. It will be also helpful for formulating scenarios related to long-term evolution in the surface environment required for performance and safety assessments of the deep geological disposal.

Many countries plan to dispose of spent nuclear fuel through deep geological disposal system. In Korea, a plan is being established for the construction of a deep disposal facility to dispose of highlevel radioactive waste (or spent nuclear fuel). For construction of a deep geological repository, the NSSC (Nuclear Safety and Security Commission) stipulate that detailed technical standards for location, structure, and disposal system of deep geological repository are determined and announced by the Nuclear Safety and Security Commission Notification. Therefore, the regulatory body should carry out the process of regulatory review whether the technical standards developed by the implementer are suitable for the IAEA’s recommendations and guidelines and domestic conditions. In this process, there are many difficulties and uncertainties in terms of time and cost to independently develop safety factors in Korea by referring to the IAEA reports. So, this study intends to investigate and analyze regulatory cases for important safety factors through cases of overseas leading countries in deep geological disposal project. There are two regulatory cases intensively investigated in this study. The first is a regulatory case of regulatory bodies and external experts on the safety case, and the second is a regulatory review case in the process of site selection factor selection. In case of regulatory review of safety case, Sweden and France were selected as the representative target countries. In Sweden, safety cases such as SR-97, SR-Can, and SR-Site have been developed and there are cases of active regulatory review by regulatory agencies in the RD&D process. In France, several safety cases based on sedimentary rocks were developed and the OECD/NEA IRT (International Review Team) was inquired for review for each safety case. The site selection process is divided into a preliminary site selection stage, a site investigation stage, and a site selection and application stage. In each stage, evaluation to select a safe site is carried out using allocated siting factors of that stage. The IAEA SSG-14 report describes aspects that implementers consider in the site selection process and, with this reference, many countries are developing various siting factors and assessment methodologies in consideration of their domestic bedrock condition and geological positions. As a representative example, in Japan which is highly affected by earthquakes and igneous activities, the siting factor is classified into EF (Evaluation Factors) and FF (Favoulable Factors). So, site assessment is conducted preferentially using EF related to earthquakes and igneous activity.

Through constructing statistical fracture network model based on discrete element method, the evolution characteristics of the fracture aperture had been directly simulated and evaluated caused by redistributed stress after the borehole excavation. This study focuses on the size effect of the discrete element method for the analysis of the effective distance of fracture aperture change after the borehole excavation. A two-dimensional trace-type domain with a maximum size of 1.1 m2 was created using a discrete fracture network with stochastic information of KURT. A total of eight domains with different sizes were constructed from the largest domain area to the 0.4 m2 analysis area. The aperture change ratio which can be depending on the domain size was examined. The ratio was investigated by comparing the aperture size before and after the simulation of borehole excavation. In addition, the effective range of aperture changes was analyzed by comparing the re-distribution distance from the center of the borehole. Based on dimensional analysis, input variables (borehole radius, occurrence distance of aperture changes, domain size) were modeled using exponential distribution form. Through the analysis model, two dimensionless variables were derived to investigate the expected distance of the aperture changes and appropriate DFN domain size for simulating bole excavation. As an application example of the 3-inch borehole simulation, the analysis model predicted that the range of aperture changes could occur within a radius of about 0.98 m from the borehole center, and the suitable size of the model had been inferred as about 5 × 5 m for minimizing the domain size effect.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

Deep geologic repositories (DGR) are designed to store spent nuclear fuel and to isolate it from the biosphere for an extended period of time as long as millions of years. The long-term performance of the DGR replies on the performance of the natural geologic barriers after the end of the lifetime for the engineered barrier systems. Typically, multiple analytical and numerical models are used to analyze and ensure the safety of the repositories along both engineered and natural barrier systems. Despite the immense advancement in computing power and modeling techniques over the last few decades, a series of models and their linkage often require many simplifying assumptions in this safety assessment. The degree of the reliability and confidence of the safety analysis is thus highly dependent on the validity of those tools used. Considering the significance of the DGR performance and public attention, the highest level of quality control is necessary for the models employed in the assessment. The performance of the ultimate long-term geologic barrier is determined by the expected travel time of the radioactive species of interest, the level of their dilution or radioactivity at compliance areas, and the uncertainty associated with them. As the species of interest can be carried away from the repository location by groundwater flow, the travel time is determined by groundwater velocity along the flow path from source to biosphere while the dilution is a function of the decay and production rates as well as the diffusion and dispersion. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the models used for safety evaluation will need to become more and more comprehensive, robust, and efficient which is difficult to achieve in principle. They will also need to be transparent and flexible to satisfy the regulatory quality control requirements. This study thus attempts to develop an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be widely accepted by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM can be considered as a tool and a framework at the same time when it is designed to 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 simple illustrative examples.

The radioactive waste repository consists of an engineered barrier and a natural barrier and must be managed safely after isolation. We classify the geological events of natural barriers for the evaluation of their present and future disposal stability assessment, they can be divided into regional and regional evolutions according to their scale. Regional evolution can be quantitatively explained by plate tectonics and regional rock distribution, and local evolution can be explained by petrological, mineralogical evidence and ductile, brittle deformation. Plate tectonics can explain the change quantitatively by restoring the direction of the Earth’s magnetic field recorded when rocks were formed. The time units for these changes are tens of millions of years to hundreds of millions of years, but plate tectonic is a way to estimate geological history. It can be assessed by extrapolating past knowledge considering the known geological events of radioactive waste repository. It is possible to derive a conservative value of the change of the geological environment in the time unit of disposal stability. The Korean Peninsula belongs to the edge of the Eurasian plate and is divided into Gyeonggi, Yeongnam Massif, Okcheon orogeny belt, and Gyeongsang Basin. To quantitatively determine their geological history, we collected paleomagnetic data using rocks from the Korea Peninsula (paleomagnetic database and papers). We attempted to carry out the apparent polar wander paths (APWPs) on the Korean Peninsula by collecting and sorting data. Since the Korean Peninsula is composed of multiple massifs, this APWP is expected to serve as a basis for explaining the local crustal rotation or brittle ductile deformation. Furthermore, by extrapolating the change pattern from the past to the present, it can contribute to the estimation of the future geological evolution.

PWR spent nuclear fuel generally showed an oxide film thickness of 100 um or more with a combustion rate of 45 MWD/MTU or higher, while CANDU spent nuclear fuel with an average combustion rate of about 7.8 MWD/MTU had few issues related to hydride corrosion. Even based on the actual power plant data, it is known that the thickness of the oxide film is 10 μm or less on the surface of the coating tube, and brittleness caused by hydride is shown from the thickness of the oxide film of about 80 μm, so it is not worth considering. However, since corrosion may be accelerated by lithium ions, lithium ions may be said to be a very important factor in controlling the hydro-chemical environment of heavy water. Lithium has a negative effect on the corrosion of zirconium alloys. However, since local below 5 ppb to prevent corrosion. maintained at a concentration between 0.35 and 0.55 ppm. Hydrogen is known to have a positive effect by suppressing radioactive decomposition of the coolant and suppressing cracks in nickelbased alloys. However, too much hydrogen can produce hydride in a pressure tube composed of Zr-2.5Nb, so DH (Disolved Hydrogen) maintains the range of 0.27–0.90 ppm. pH and conductivity are completely determined by lithium ions, and DH can be completely removed below 5 ppb to prevent corrosion. Therefore, for cladding corrosion simulation of the CANDU spent nuclear fuel, a hydrochemical of the equipment, not 310°C, and 14 uS·Cm−1 is targeted as conditions for corrosion acceleration. In addition, for acceleration, the temperature was set to 345°C (margin 10°C), which is the maximum accommodation range of the equipment, not 310°C.

When disposing of spent nuclear fuel, there is a risk of exposure that could exceed the annual allowable dose due to human intrusion after the institutional control period. Therefore, it can be treated with the pyroprocess, but the decontamination factor is not sufficient, and an additional actinide recovery is required because molten waste salt-containing actinide is generated. In the case of reducing the element in the spent molten salt through an electrochemical method using a liquid Bi electrode, it is difficult to separate only the actinide element because the two-element groups are reduced together due to the large concentration difference between the actinide and the rare earth element. Therefore, a process of forming a Bi intermetallic compound using a liquid Bi electrode, which has higher element separation efficiency than a liquid Cd electrode, and physically separating the Bi intermetallic compound using the difference in density of the produced compound has been proposed. For this, it is necessary to understand the properties and density separation of the intermetallic compound to be produced, and experiments were planned and conducted for this purpose. Various metals were added to the molten Bi to form an intermetallic compound, and an analysis device such as SEM was used to determine the intermetallics distribution, composition, and internal structure. As the added metal, Ce is a representative element for lanthanide, and Hf with the most similar intermetallic density, decomposition temperature, and standard reduction potential to U, and U as a substitute element for actinide was adopted. As a result of SEM and EDS analysis, it was confirmed that the separation was made in Bi due to the density difference between the produced intermetallic compounds. A Ce-Bi intermetallic compound was observed in the upper part, Hf at a concentration smaller than the error range was detected, and a Hf-Bi intermetallic compound which containing high concentration of Ce was observed in the lower part. Separation of high-purity Ce seems to be possible in the upper part, and it seems difficult to separate high-purity Hf in the lower part. Therefore, to separate highpurity Hf, an additional process suitable for it seems to be necessary.

The fabrication of waste forms with high thermal and structural stability is an essential technology for the safe management and disposal of radioactive wastes. In particular, the thermal characteristics of waste forms containing high heat-generating nuclides such as Cs and Sr can be used for the optimized design of the waste form to secure its thermal safety, and they also provide basic design data for the safe management of canisters, storage systems, and repositories. The Korea Atomic Energy Research Institute is actively developing processes and equipment for fabricating various types of high-level wastes into a stable glass or ceramic waste form. In previous research related to the thermal analysis of the waste form, a relatively simple analysis was performed by using the analytic solution of the one-dimensional steady-state heat conduction equation considering the decay heat properties of the waste. As a specific application study, the optimized diameter of the cylindrical glass waste form was proposed by evaluating the centerline temperature of the waste form. In this study, we extended previous research by introducing a more complicated model, and the main results are summarized as follows. First, an analytical solution was derived by applying the temperaturedependent thermal conductivity expressed in the general form of polynomial function to the onedimensional heat conduction problem previously studied. Second, the two-dimensional axisymmetric steady-state heat conduction problem with a more realistic cylinder model with finite length was modeled and solved by using the finite element method via Matlab’s PDE (partial differential equation) toolbox. Third, thermal analysis was performed on the SrTiO3 waste form, selected as a stable form of strontium nuclide, using the developed analytical and numerical methods. The differences in the temperature distribution and computation time were evaluated through a comparative study of both solutions. Although the problem considered in this study could easily be solved by using commercial CFD software such as ANSYS or SolidWorks, a code-based program was developed to facilitate parametric design study in conjunction with optimization algorithms. The analysis results could be used to evaluate the thermal stability of waste form and to optimize the shape and size of the waste form in consideration of the design constraints of storage systems or repositories.

이 논문의 목적은 악셀 호네트의 인정투쟁이론과 칼 슈미트의 정치개 념으로부터 발전시킨 ‘적대와 인정의 정치’ 틀로 공산체제 이후 헝가리 정치변화를 고찰하는 데 있다. 헝가리에서는 1990-2010년의 체제전환 기간에 온건다당제와 양당제를 가진 비교적 공고한 민주주의가 자리잡고 인정의 정치가 유지되었다. 정치세력 사이에 경쟁과 갈등은 존재했지만, 상대방을 정치무대에서 제거하고자 하는 적대의 정치를 발견하기는 어려 웠다. 그러나 2010년 이래 헝가리의 정치는 엄청나게 변모하였다. 2010 년에 복귀한 오르반정부는 포퓰리즘을 대변하였고 3연임에 성공하였다. 오르반정부의 포퓰리즘은 비자유적 민주주의로 정당화되어왔다. 그의 정 부는 민주세력이나 반대세력을 억압하였고 언론자유를 침해해왔다. 또한 민족주의적 감정을 동원하거나 조장하였고, 반EU정책이나 반난민정책을 전개하였다. 오르반정부의 포퓰리즘으로 인해 헝가리에서는 인정의 정치 가 적대의 정치로 변모한 것이다. 2010년 이후 헝가리 민주주의는 후퇴 하고 악화되었으며, 헝가리정치는 민주주의와 권위주의 사이에서 표류하 고 있다. 지오반니 사르토리의 정당체제이론을 헝가리에 적용하면, 헝가 리는 일당우위치제와 패권정당체제의 경계선에 위치한 것으로 보인다.