Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

The mobility of radionuclides in the subsurface environment is governed by a interaction of radioactivity characteristics and geochemical conditions with adsorption reactions playing a critical role. This study investigates the characteristics and mechanisms of radionuclides adsorption on site media in viewpoint of nuclear safety, particularly focusing on the potential effect of seawater infiltration in coastal site near nuclear power plant. Seawater intrusion alters the chemistry in groundwater, including parameters such as pH, redox potential, and ionic strength, thereby affecting the behavior of radionuclides. To assess the safety of site near nuclear power plant and the environmental implications of nuclide leakage, this research conducted various experiments to evaluate the behavior of radionuclides in the subsurface environment. High distribution coefficients (50-2,500 ml/g) were observed at 10 mg/L Co, with montmorillonite > hydrobiotite > illite > kaolinite. It decreased with competing cations (Ca2+) and was found to decrease significantly by 90% with a decrease in pH to 4. It is believed that the adsorption capacity of cationic radionuclides decreases significantly as the clay mineral surface becomes less negatively charged. For Cs, the distribution coefficient (180-560 ml/g) was higher for montmorillonite > hydrobiotite > illite > kaolinite. Compared to Co, it was found to be less influenced by pH and more influenced by competing cations. For Sr, the distribution coefficient (100-380 ml/g) was higher in the order of hydrobiotite > montmorillonite > illite > kaolinite. Compared to Cs, it was found to be less affected by pH and also less affected by the effect of competing cations compared to Cs. Seawater samples from Gampo and Uljin site near Nuclear Power Plant in Korea were analyzed to determine their chemical composition, which was subsequently used in adsorption experiments. Additionally, the seawater-infiltrated groundwater was synthesized in laboratory according to previous literature. The study focused on the adsorption and behavior of three key radionuclides such as cesium, strontium, and cobalt onto four low permeability media (clay minerals) such as kaolinite, illite, hydrobiotite, and montmorillonite known for their high adsorption capacity at a site of nuclear power plant. At concentrations of 5 and 10 mg/L, the adsorption coefficients followed the order of cobalt > cesium > strontium for each radionuclide. Notably, the distribution coefficient (Kd) values exhibited higher values in seawater-infiltrated groundwater environments compared to seawater with relatively high ionic strength. Cobalt exhibited a substantial adsorption coefficient, with a marked decrease in Kd values in seawater conditions due to elevated ionic strength. In contrast, cesium displayed less dependency on seawater compared to other radionuclides, suggesting distinct adsorption mechanisms, possibly involving fractured edge sites (FES) in clay. Strontium exhibited a significant reduction in adsorption in seawater compared to groundwater in all Kd sorption experiments. The adsorption data of cobalt, cesium, and strontium on clay minerals in contact with seawater and seawater-infiltrated solutions offer valuable insights for assessing radioactive contamination of groundwater beneath coastal site near nuclear power plant sites. This research provides a foundation for enhancing the safety assessment protocols of nuclear power plant sites, considering the potential effects of seawater infiltration on radionuclide behavior in the subsurface environment.

When dismantling a power plant, a large amount of radioactive tanks are generated, and it is estimated that a significant amount of sludge will accumulate inside the tanks during long-term operation. In the process of dismantling a radioactive tanks, it is important to know the composition of the sludge because the sludge present inside must first be removed and then disposed of. In the case of certain tanks, it can be predicted that corrosion products generated due to system corrosion are the main cause of sludge formation. However, in the case of some tanks, it is not easy to predict the sludge composition because various dispersed particles in addition to corrosion products may be mixed with the wastewater. Even if it is collected and analyzed, the sludge composition can change significantly depending on the operation history, so the analysis results cannot be considered representative of the composition. In the case of LHST, surfactant components introduced during the washing and shower process, oil components and dispersed particles dissolved by the surfactant accumulate inside the tank, making sludge difficult to remove. In addition, even if it is removed by ultra-high pressure spraying, unexpected problems may occur in the subsequent treatment process due to the surfactant contained therein. Therefore, it is necessary to analyze in more detail the characteristics of sludge accumulated in LHST and prepare countermeasures. A test procedure was prepared to evaluate the characteristics of sludge accumulating in LHST. According to the test results, the long-term sludge accumulation tendency of the LHST is summarized as follows. ① Initially, the sludge settling speed increases slowly until a surface sludge layer is formed. ② After the surface sludge layer is formed, the sludge rapidly settles until the sludge layer becomes somewhat thicker. ③ When the sludge layer is formed to a certain extent, the sludge escape rate increases and the sludge accumulation rate decreases again. It is assumed that the sludge escape speed is closely related to the fluid flow speed in the relevant area. It is believed that the combined effect of these phenomena will determine the thickness of the sludge layer that will accumulate inside the tank, but it was not possible to evaluate how much the sludge layer would accumulate based on the experimental results alone. However, it can be assumed that significant sludge accumulation occurred in areas where fluid flow was minimal and sludge formation nuclei easily accumulates.

After the Fukushima accident in 2011, relevant concerns regarding the contamination of the natural environment rose abruptly. For example, water contaminated by radionuclides such as Cs and Sr may directly flow into the ocean and threaten the marine ecosystem. In this respect, costeffective and efficient decontamination techniques need to be developed and verified to remediate the contaminated water. Prussian blue (PB) is known as a representative material that can adsorb Cs by ion-trapping and is widely used for medical purposes. However, there is a limitation that PB itself is non-separable and highly mobile in aqueous system, so it needs a fixture, such as bentonite, to be collected after the adsorption. Furthermore, while the performance of PB toward Cs is relatively well known, its behavior toward Sr has rarely been reported. The object of this study is to investigate the sorption characteristics of Cs and Sr onto PB-functionalized bentonite at various conditions. The adsorbent employed in the present work was prepared by mixing bentonite, FeCl3, and K4[Fe(CN)6] at room temperature for 24 hours in the aqueous solution. The concentrations of FeCl3 and K4[Fe(CN)6] were set to a range of 5-200 % compared to the cation exchange capacity of bentonite. After that, the PB-functionalized bentonite was sieved with a mesh size of 63 μm and then reacted with the Cs and Sr solution at various liquid-to-solid (L/S) ratios of 2-10 g/L for up to 500 minutes. Moreover, synthetic seawater containing additional Cs and Sr was reacted with PBfunctionalized bentonite to characterize the ion selectivity of PB. After the completion of the adsorption experiment, a part of the adsorbent was separated and desorption of Cs and Sr with 2 M of nitric acid was performed. For the quantification of aqueous Cs and Sr concentrations, ICP-MS was employed after the filtration with a pore size of 0.45 μm. The result obtained in this study revealed a high sorption affinity of Cs and Sr onto PBfunctionalized bentonite. The analysis results also presented that the sorption reactions of Cs and Sr reached their steady state within 10 minutes of reaction time. Furthermore, the ion selectivity toward Cs and Sr was verified through sorption test with synthetic seawater. According to the high sorption affinity and selectivity, the PB-functionalized bentonite synthesized through this study is expected to be widely used for remediating the Cs- and Sr-contaminated groundwater and seawater, particularly in nuclear waste-relevant industries.

Wasteform is the first barrier to prevent radionuclide release from repositories into the biosphere. Since leaching rates of nuclides in wasteform significantly impact on safety assessment of the repository, clarifying the leaching behavior is critical for accurate safety assessment. However, the current waste acceptance criteria (WAC) of the Gyeongju repository only evaluates leachability indexes for Cs, Sr, and Co, which are tracers for nuclear power plant waste streams. Furthermore, ANS 16.1, the current leaching test method used in WAC, applies deionized water (DI) as leachant. However, the interactions between wasteform and groundwater environment in the repository may not be reflected. Therefore, it is necessary to review the current leaching test method and nuclides that may require the extra evaluation of leachability beyond the Cs, Sr, and Co. Tc and I are key nuclides contributing to high radioactive dose in safety assessment due to their high mobility and low retardation factor. The groundwater conditions within the repository, such as pH and Eh significantly affect the chemical form of Tc and I. For example, Tc in H2O system tends to form hydroxide precipitates in neutral pH condition and TcO4 - in strong alkaline environments according to the Pourbaix diagram. In case of I, it generally exists in the form of I-, while it exists as IO3 - as Eh increases. Although the current leaching test at the Gyeongju repository applies DI as a leachant, the actual repository is expected to have a highly alkaline environment with a substantial amount of various ions in the groundwater. Consequently, the leaching behavior in the ANS 16.1 test and the actual disposal condition is different. Thus, it is necessary to analyze the leaching behavior of Tc and I with reflecting the actual disposal environment. In this study, the leaching behavior of Tc and I is investigated by following ANS 16.1 leaching test method. The solidified waste specimens containing 10 mmol of Re and I were manufactured with cement, which is widely used as a solidification material. Re was applied instead of Tc, which has similar chemical behavior to Tc, and NH4ReO4 and NaI were used as surrogates for Re and I. As a leachant, deionized water and cement-saturated groundwater were prepared and the concentration of nuclides in the leachant is analyzed by ICP-OES. As the result of this study, experimental data can be applied to improve the WAC and disposal concentration standards in the future.

The buffer is installed around the disposal canister, subjected to heating due to decay heat while simultaneously experiencing expansion influenced by groundwater inflow from the surrounding rock. The engineering barrier system for deep geological disposal require the evaluation of longterm evolution based on the verification of individual component performance and the interactions among components within the disposal environment. Thus, it is crucial to identify the thermalhydro- mechanical-chemical (THMC) processes of the buffer and assess its long- and short-term stability based on these interactions. Therefore, we conducted experimental evaluations of saturationswelling, dry heating, gas transport, and mineralogical alterations that the buffer may undergo in the heated-hydration environment. We simulated a 310 mm-thick buffer material in a cylindrical form, simulating the domestic disposal system concept of KRS+ (the improved KAERI reference disposal system for spent nuclear fuel), and subjected it to the disposal environment using heating cartridges and a hydration system. To monitor the thermal-hydro-mechanical behavior within the buffer material, load cells were installed in the hydration section, and both of thermal couples and relative humidity sensors were placed at regular intervals from the heat source. After 140 days of heating and hydration, we dismantled the experimental cell and conducted post-mortem analyses of the samples. In this post-mortem analysis, we performed functions of distance from the water contents, heat source, wet density, dry density, saturation, and X-Ray diffraction analysis (XRD). The results showed that after 140 days in the heated-hydration environment, the samples exhibited a significant decrease water contents and saturation near the heat source, along with very low wet and dry densities. XRD Quantitative Analysis did not indicate mineralogical changes. The findings from this study are expected to be useful for input parameters and THMC interaction assessments for the long-term stability evaluation of buffer in deep geological disposal.

The presence of technological voids in deep geological repositories for high-level radioactive nuclear waste can have negative effects on the hydro-mechanical properties of the engineered barrier system when groundwater infiltrates from the surrounding rock. This study conducted hydration tests along with image acquisition and X-ray CT analysis on compacted Korean bentonite samples, which simulated technological voids filling to investigate the behavior of fracturing (piping erosion) and cracking deterioration. We utilized a dual syringe pump to inject water into a cell consisting of a bentonite block and technological voids at a consistent flow rate. The results showed that water inflow to fill technological voids led to partial hydration and self-sealing, followed by the formation of an erosional piping channel along the wetting front. After the piping channel generated, the cyclic filling-piping stage is characterized by the repetitive accumulation and drop of water pressure, accompanied by the opening and closing of piping channels. The stoppage of water inflow leads to the formation of macro- and micro cracks in bentonite due to moisture migration caused by high suction pressure. These cracks create preferential flow paths that promote longterm groundwater infiltration. The experimental test and analysis are currently ongoing. Further experiments will be conducted to investigate the effects of different dry density in bentonite, flow rate, and chemical composition of injected water.

Recently, as carbon-neutral energy sources become increasingly important worldwide, SMRs (Small Modular Reactors), which offer significantly enhanced safety, versatility, and mobility compared to conventional nuclear reactors, are gaining attention as a viable alternative. SMR generally refers to small modular reactors with a power output of 300 MWe or less. Unlike conventional reactors, SMRs are characterized by an all-in-one design where peripheral systems and equipment are all integrated into the reactor itself, leading to enhanced reliability and durability. Additionally, the nuclear fuel reloading cycle is significantly extended compared to traditional reactors, resulting in a substantial reduction in maintenance difficulty and costs. Researchers have taken note of these characteristics of SMRs, particularly the extended fuel reloading cycle. Therefore, we have initiated the initial design of an ultra-small Micro Modular Reactor with an electricity generation capacity of 10 MWe and a fuel cycle of up to 55 years, with the goal of using it as a propulsion power source for various transportation modes, especially ships. Our design of MMR, called ‘ARA,’ is primarily distinguished by its use of U233 and Th232 fuels instead of conventional UO2 fuel. Due to various features of ‘ARA,’ including different fuel compositions, ARA is predicted to exhibit several characteristic features compared to conventional PWRs. In this study, among these characteristics, we focused on predicting changes in material composition within the fuel rod during the extended cycle operation of high-enriched fuel, rather than short-cycle operation using low-enriched fuel, unlike conventional reactors. The primary goal of this research is to observe the behavior of the composition of the materials used in the fuel cycle of the MMR, which utilizes U233 and Th232 fuels instead of UO2. Considering the difficulties in the spent nuclear fuel disposal process, many different trials were made to minimize the fission products of ARA, which differs from conventional reactors in terms of fuel type, size, and fuel cycle, in relation to waste generation.

To investigate the mechanical integrity of spent nuclear fuel, the failure behavior of the cladding tube was examined under accident conditions. According to the SNL report, the failure behavior of cladding can be broadly classified into two types. The first is failure due to bending load caused by falling. The second is failure due to pinch load caused by space grid. In this study, mechanical integrity was evaluated through the stress intensity factor applied to the crack in failure behavior due to bending load. Since the exact value of the impact load due to fall was unknown, the load was applied by increasing the value up to 200 G in 20 G increments. The size of the crack is an important input variable, and 300 um was given by referring to the EPRI report, and the elastic modulus, a material property that determines the stress field, was given 75.22 GPa by referring to the FRAPCON code. Since the relationship between the direction of stress and the direction of the crack is also a major variable, simulations were conducted for both cracks perpendicular to and parallel to the stress direction. It was confirmed that at a load of 200 G, when the crack was parallel to the stress direction, stress concentration did not occur and had a very low stress intensity factor 0.01 􀜯􀜲􀜽√􀝉. When perpendicular to the direction of stress, the stress intensity factor showed a value of 1 􀜯􀜲􀜽√􀝉. However, considering that the critical value of the stress intensity factor due to hydride is 5 􀜯􀜲􀜽√􀝉, it can be seen that perpendicular result also ensures the mechanical integrity of the cladding.

Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

Zircaloy-4 is utillzed in nuclear fuel rod cladding due to its strength and corrosion resistance. However, it can undergo deformation over time, known as creep, which poses a safety risk in reactors. Furthermore, hydrogen absorption during reactor operation can alter its properties and affect creep rates. Previous research suggests a trend in which hydrogen concentration corelates unidirectionally with creep rates, either increasing or decreasing as the concentration rises. This trend can also be observed in EPRI’s creep model, EDF-CEA Model-3. However, recent literature has suggested that creep behavior may vary depending on the state of hydrogen presence. Therefore, it has become evident that creep behavior can be influenced not only by hydrogen concentration but also by the state of hydrogen presence, whether it is in a solid solution state or precipitated as hydrides. Our study aimed to compare creep behavior in specimens with hydrogen concentrations below and above solubility limits. We fabricated Zircaloy-4 plate specimens with varying hydrogen concentrations and conducted creep tests. The results revealed that specimens below the solubility limit exhibited decreasing creep rates as hydrogen concentration increased, while those above the limit displayed increasing creep rates. This investigation confirms that the state of hydrogen presence significantly impacts creep behavior within Zircaloy-4 cladding. As part of our additional research plans, we intend to conduct creep tests on the material based on its orientation, whether it is in the rolling direction (RD) or the transverse direction (TD). We also plan to perform creep tests on ring specimens. Additionally, for the ring specimens, we aim to evaluate how creep behavior differs between the cold-worked stress-relieved (CWSR) condition and the recrystallized annealed (RXA) condition achieved through high-temperature heat treatment.

The demand for high-strength steel is rising due to its economic efficiency. Low-cycle fatigue (LCF) tests have been conducted to investigate the nonlinear behaviors of high-strength steel. Accurate material models must be used to obtain reliable results on seismic performance evaluation using numerical analyses. This study uses the combined hardening model to simulate the LCF behavior of high-strength steel. However, it is challenging and complex to determine material model parameters for specific high-strength steel because a highly nonlinear equation is used in the model, and several parameters need to be resolved. This study used the particle swarm algorithm (PSO) to determine the model parameters based on the LCF test data of HSA 650 steel. It is shown that the model with parameter values selected from the PSO accurately simulates the measured LCF curves.

본 연구는 간호대학생의 건강관심도와 건강증진행위와의 관계에서 건강관리 자기효능감의 매개 효과를 규명하기 위해 시행되었다. 연구 대상자는 G시와 J도에 소재한 4개 대학교의 간호학과 재학생 255 명이다. 수집된 자료는 SPSS 25.0프로그램을 이용하여 기술통계와 t-test, ANOVA, Pearson correlation, 회귀분석을 시행하였으며 매개효과 검증을 위해 Sobel test를 실시하였다. 연구결과, 건강관리 자기효능감 은 건강관심도와 건강증진행위 사이에서 부분매개효과를 나타냈고(Z=5.54, p<.001), 건강관심도와(β =.20, p<.001) 건강관리 자기효능감이(β=.39, p<.001) 높을수록 건강증진행위 수준이 높아지며, 건강증진 행위를 설명하는 설명력(R2)은 48%이었다. 따라서 간호대학생의 건강증진행위를 향상시키기 위해서는 건 강관심도를 높일 뿐만 아니라 건강관리 자기효능감을 동시에 강화시킬 수 있는 효과적인 전략 마련이 필요 하다.

이 연구는 핵무기 폭발 시 발생하는 효과 변인을 토대로 북한이 언제, 어떤 방법으로 핵무기를 운용할 것이며, 핵폭발 시 생성되는 방사성 물 질이 자연환경과 인공물의 영향에 따라 도심지에서 어떤 거동 현상을 보 이는가와 이를 고려한 국민 방호의 대비 방향에 관한 것이다. 연구 결과 핵무기는 폭발 고도에 따라 그 효과가 달라지며, 북한은 이를 활용하여 개전 초부터 가장 효과적인 공격을 할 것으로 예측되었다. 즉, 북한은 개 전 초 한미연합군과 정부의 지휘‧통제‧통신체계를 무력화하기 위해 저위 력핵무기로 지하 폭발을, 전쟁 도중 결정적인 목표 확보를 위해 전술핵 무기로 저공 폭발을, 전쟁 말기 패색이 짙어지는 위기 시에는 전술핵무 기로 지표면 폭발을 시도할 것이다. 북한의 핵무기 공격 후 발생되는 방 사성 물질은 낙진의 형태로 일정 지역을 오염시킬 것이며, 방사성 물질 이 도심지로 유입된다면 공기역학 또는 유체역학적 거동을 보임으로써 다양한 형태의 오염과 위험이 존재할 것으로 분석되었다. 이에 따라 국 민 방호를 위해서는 북한의 핵무기 공격 양상을 고려 최악의 상황을 가 정한 대비가 평시에 완료되어야 하며, 전쟁 개시 이후에는 당시의 공격 유형에 부합한 대응 및 복구가 뒤따라야 한다. 아울러 방사능 낙진의 거 동을 세밀히 분석하고 이를 고려하여 핵폭발 초기 효과에 대비하는 주민 대피와 이를 후속하는 낙진에 대응하기 위한 주민 소개는 분리되어야 한다.

Molybdenum-tungsten (Mo-W) alloy sputtering targets are widely utilized in fields like electronics, nanotechnology, sensors, and as gate electrodes for TFT-LCDs, owing to their superior properties such as hightemperature stability, low thermal expansion coefficient, electrical conductivity, and corrosion resistance. To achieve optimal performance in application, these targets’ purity, relative density, and grain size of these targets must be carefully controlled. We utilized nanopowders, prepared via the Pechini method, to obtain uniform and fine powders, then carried out spark plasma sintering (SPS) to densify these powders. Our studies revealed that the sintered compacts made from these nanopowders exhibited outstanding features, such as a high relative density of more than 99%, consistent grain size of 3.43 μm, and shape, absence of preferred orientation.

With the increasing attention to environmental pollution caused by particulate matter globally, the automotive industry has also become increasingly interested in particulate matter, especially particulate matter generated by automobile brake systems. Here, we designed a coating composition and analyzed its mechanical properties to reduce particulate matter generated by brake systems during braking of vehicles. We designed a composition to check the mechanical properties change by adding Cr3C2 and YSZ to the WC-Ni-Cr composite composition. Based on the designed composition, coating samples were manufactured, and the coating properties were analyzed by Vickers hardness and ball-on-disk tests. As a result of the experiments, we found that the hardness and friction coefficient of the coating increased as the amount of Cr3C2 added decreased. Furthermore, we found that the hardness of the coating layer decreased when YSZ was added at 20vol%, but the friction coefficient was higher than the composition with Cr3C2 addition.

Considering that the number of middle-aged single-person households is increasing, this study investigates dietary behaviors, nutrient intake, and mental health according to household type. Data were procured from the 2015-2019 Korea National Health and Nutrition Examination Survey (KNHANES). Totally, 5,466 participants aged 50-64 years were classified into 2 groups: a household with one member was defined as a single-person household, and households with two or more members were described as multi-person households. Single-person households comprised 10.63% of the total, with a higher average age, and lower income and economic levels than multi-person households. Compared to multiperson households, single-person households had a higher frequency of skipping breakfast, eating alone, and dining out, the moderately and severely food insecure group was more than 5 times, and nutrient intake and dietary quality were poorer. In the fully adjusted model, the odds ratios (ORs) of depressive symptoms were 2.35 times (95% CI: 1.39-3.96), and suicide ideation was 1.95 times (95% CI: 1.35-2.82) in single-person compared to multi-person households. Our results lead us to conclude that poor dietary intake in middle-aged single-person households affects the mental health, and the above factors should be considered when framing the dietary policy.

PURPOSES : This study aims to understand the effect of defective dowel bar installation on jointed concrete pavement (JCP), which can cause joint freezing, spalling, cracking, and faulting and finally shorten the lifespan of the pavement. METHODS : A comprehensive field survey was undertaken at an expressway construction site in South Korea to assess dowel bar installation conditions. In addition, finite element (FE) analysis was used to simulate JCP behavior with both vertical and horizontal dowel misalignments. Different temperature conditions, including a change of -55 °C and gradient of -0.1 °C/mm, were integrated into the FE model to examine horizontal slab contraction and simultaneous slab curling. RESULTS : The analysis revealed pronounced slab behaviors under specific temperature changes, particularly when combined with dowel misalignments. The simultaneous effects of horizontal contraction and slab curling owing to temperature changes and gradients became more evident in the presence of dowel misalignments. CONCLUSIONS : The results confirmed that dowel bar misalignment considerably affected the behavior of the JCP, thereby emphasizing the importance of the proper installation of dowel bars.

PURPOSES : In this study, the initial behaviors of shoulder concrete slabs and tiebars at the longitudinal construction joints between the shoulder JCP and mainline CRCP are investigated. METHODS : The strains of concrete and tiebars were measured at the longitudinal construction joint between the CRCP and JCP. Measurements were performed using data accumulated over a month after concrete placement. The contact conditions were investigated by comparing the strains at each location. RESULTS : The longitudinal construction joints between the shoulder JCP and mainline CRCP exhibited the composite behavior of bonding and friction, and a virtual neutral axis was formed inside the JCP. At the connection of the shoulder concrete, the strain and temperature of the concrete were inversely proportional. The tiebars connecting the CRCP and JCP exhibited different behaviors depending on the bonding conditions around the tiebars of the construction joints. In the presumed state in which the bonding condition was maintained, the concrete temperature and tiebar strain were directly proportional; however, the presumed state of the separation condition exhibited an inversely proportional relationship. In the 24-h behaviors of the tiebars, the effects of the horizontal and curling movements overlapped, and the strains of the measured tiebars increased at the minimum and maximum temperatures of the shoulder JCP. CONCLUSIONS : The strains in the tiebars and concrete slabs primarily depended on the boundary conditions (bonding and friction) of the longitudinal construction joint between the shoulder JCP and mainline CRCP.