본 논문은 분점정부 하에서 국회 효율성을 ‘공공선택론(Public Choice)’ 의 관점에서 분석하였다. 기계·경제적, 사회·정치적 등 다양하게 측정되 는 효율성의 개념을 살펴보고, 국회 효율성을 국회 위원회 제도, 대통령 거부권, 국회의원의 활동 및 국회 의안 처리를 기준으로 분석하였다. 구 체적으로 뷰캐넌과 톨록의 집단의사결정에 관한 이론을 근거로 국회의 위원회 중심주의 운영제도가 어떻게 효율성을 제공하는지를 분석하였다. 또한 거대 양당의 대결 구도가 더 심각해지는 분점정부(여소야대) 구조가 국회 효율성에 미치는 영향을 대통령 거부권 제도와 연계하여 분석하였 다. 본 연구는 제21대 국회 상반기는 단점정부로, 하반기는 분점정부 구 조라는 데 주목하면서, 대통령 거부권 행사, 국회의원의 활동, 국회의 의 안 처리 등을 각각 비교·분석하였다. 제21대 국회의 효율성은 단점 정부 인 상반기가 하반기보다 상대적으로 높게 나타났다. 분점정부에서 나타 날 수도 있는 국회 효율성이 상쇄된 것은 여야의 양당 정쟁 구조가 대통 령과 국회의 극단적인 대결 구도로 굳어지고 있다는 것을 시사한다.

Seawater evaporation and purification powered by solar energy are considered as a promising approach to alleviate the global freshwater crisis, and the development of photothermal materials with high efficiency is imminent. In this study, cellulose nanofiber (CNF)/MXene/Ni chain (CMN) aerogels were successfully synthesized by electrostatic force and hydrogen bond interaction force. CMN10 achieved a favorable evaporation rate as high as 1.85 kg m− 2 h− 1 in pure water, and the corresponding evaporation efficiency could be up to 96.04%. Even if it is applied to seawater with multiple interference factors, its evaporation rate can still be 1.81 kg m− 2 h− 1. The superior seawater evaporation activity origins from the promoted separation of photoexcited charges and photothermal conversion by the synergy of Ni chain and MXene, as well as the water transport channel supported by the 3D structure frame of CNF. Most importantly, CMN aerogel can maintain water vapor evaporation rates above 1.73 kg m− 2 h− 1 under extreme conditions such as acidic (pH 2) and alkaline (pH 12) conditions. In addition, various major ions, heavy metals and organic pollutants in seawater can be rejected by CMN10 during desalination, and the rejection rates can reach more than 99.69%, ensuring the purity of water resources after treatment. This work shows the great potential of CMN aerogel as a high-efficiency solar evaporator and low-cost photothermal conversion material. Cellulose nanofiber (CNF)/MXene/Ni chain (CMN) aerogels demonstrated high evaporation of water from sea water.

본 연구는 역사적 제도주의를 활용해 박정희 정부의 국회의원 선거제 도를 생성 과정, 지속 과정, 조정 과정, 변화 과정으로 검토하고자 하였 으며, 그 결과는 다음과 같다. 첫째, 생성 과정이다. 권력장악을 위한 정 당추천제, 소선거구제, 제1당에 유리한 전국구 배분 원칙이 적용된 국회 의원선거법은 집권당에게 득표율 대비 의석률에 유리하였다. 둘째, 지속 과정이다. 제1당에 유리한 국회의원선거법이 유지되면서 집권당의 일당 우위정당체제는 더욱 강화되었다. 셋째, 조정 과정이다. 기존 선거제도에 서 의석수의 확대, 등록기간의 축소, 배분 제한조건의 강화 등 변화가 있 었고, 집권 세력에 대한 국민의 저항으로 제1야당은 집권당을 견제할 수 있게 되었다. 넷째, 변화 과정이다. 집권위기에 직면한 집권세력은 중선 거구제를 도입하여 지역선거구에서 국회의 과반수 의석을 확보하였다. 박정희 정부에서 제1당은 이득비와 보너스율에서 이득을, 제2당은 제9대 국회의원선거를 제외하면 이득비와 보너스율에서 손해를 보았다. 박정희 정부의 국회의원 선거제도는 제1당에 유리한 양당 체제를 강화하였으나 집권세력이 권력창출, 권력유지, 권력연장, 영구집권을 위해 국회의원 선 거제도를 도구로 활용했다는 비판에 직면하였다.

Noble metal phase, present in used fuel, are fission products that can be found as metallic precipitates in used nuclear fuel. They exist as small particles (nm~um) in grain boundaries of the used fuels. Since they are particles deposited between the grain structures, they can be considered as defects in the pellet structure. Thermal expansion of fuels with noble metal is slightly higher than that of bare fuels. The fuels at high temperature, such as immediately after being discharged from nuclear reactors, may be subject to fuel failure if sufficient cooling is not provided. Recent research has shown that the noble metals can migrate into the rim space between the pellet and the cladding, and be deposited in the inner layer of the claddings. therefore, the mechanical integrity of the cladding can be degraded by noble metals, as well as the pellets. The concentration of the noble metal phase should be considered to evaluate the effect of the noble metals on the fuel integrity, after discharge from the reactors. SCALE/ORIGEN code was used to evaluate the noble metals in fuel assembly-scale, and the radial distribution in the fuel assembly. The radial distribution of the reactor power was derived from the SCALE/TRITON, considering Westinghouse 17×17. Square cell model was chosen for the geometry and 1/4 model was applied to reduce the computation time.

A comparison and validation between the analysis and vibration test data of a nuclear fuel assembly were conducted. During the comparison and validation process, various parameters that govern the vibration behavior of the fuel assembly were determined, including nuclear fuel rod’s stiffness, spring constants of the dimple and spring of support structures, and damping coefficients. The calibration of the vibration analysis model aimed to find analysis parameters that can accurately simulate the vibration behavior of the test data. For calibration, power spectral density (PSD) diagrams were generated for both the measured signals from the test and the calculated signals from the analysis. The correlation coefficient between these two PSD plots was calculated. To find the analysis parameters, each parameter was defined as a variable with an appropriate range. Latin hypercube sampling was used to generate multiple sample points in the variable space. Analysis was performed for the generated sample points, and PSD plot correlation coefficients were calculated. Using the generated sample points and their corresponding results, a Gaussian Process Regression model was implemented for PSD plot correlation coefficients and the maximum PSD value. Based on the constructed surrogate model, the optimal analysis parameters were easily found without additional computations. Through this method, it was confirmed that the analysis model using the optimal parametes appropriately simulates the vibration behavior of the test.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.

Post Irradiation Examination Facility (PIEF) is a test facility for nuclear fuel research and development and performance evaluation. From the past to the present, assemblies and fuel rods have been transported from nuclear power plants (NPP) several times, and various destructive and non-destructive tests have been performed. Among these, in the case of the 14×14 Westinghouse STD assemblies that are transported as a whole assembly, the top nozzle is connected to the guide tube by welding. Therefore, the fuel rods could not be removed from the assembly at the NPP, so the assemblies were transported to PIEF as is. Then, after cutting between the top nozzle and the guide tube in the pool, and the fuel rods were extracted and tested. In order to transport the assembly in the future, it is necessary to maintain stability by inserting the dummy rod into the unit cell from which the fuel rod is extracted. However, since the length of the dummy rod is almost 4 m and the diameter is about 10 mm, the dummy rod often bends while passing through the dimple spring of the grid. Additionally, when dummy rods are inserted into unit cells that are continuously empty after the fuel rods are extracted, there may be cases where the dummy rods are not inserted into the desired unit cell but are bent and incorrectly inserted into the next unit cell. The moment the dummy rods are inserted into the dimple spring of grid, a load is applied to the dummy rod due to the tension of the spring. If it can be inserted while offsetting the load, the work can be performed more smoothly. Accordingly, an underwater handling tool was developed that can be inserted while offsetting the tension of the spring. Using this handling tool applies a load to the dummy rod and rotates the dummy rod itself, offsetting the tension of the spring and allowing the dummy rod to be inserted without bending. This handling tool is equipped with a shock absorbing device to protect the dummy rod and spring, and a module to rotate the dummy rod. As a result of inserting the dummy rod using the developed handling tool, it was possible to easily insert the dummy rod into unit cells that were previously impossible to insert.

Various types of spent fuel assembly in nuclear power plants have been transported to a post irradiation examination facility (PIEF) in KAERI to examine the mechanical and chemical properties of fuel and cladding. Once the fuel assembly arrive at PIEF, it is dismantled in a pool area to extract the fuel rods. Dismantling of the fuel assembly is performed by cutting the top nozzle. Currently, couple of dismantled assemblies have been stored in a storage pool without the top nozzle in PIEF. These assemblies cannot be handled directly using a gantry crane in the pool, and thus are contained in a special basket to handle. In this research, we developed a restoration method for a dismantled spent fuel assembly, especially for 16×16 Korea Optimized Fuel Assembly (KOFA). After reviewing the original design document and reports of KOFA, two tools are devised; an assembly tool and a tightening tool for a bolt. Since the top nozzle and dismantled KOFA can be re-assembled using a bolt, we follow the original design, size, and materials of the previously used bolt. The bolt to restore the top nozzle of KOFA is made of 321 stainless steel and has a design that fits the guideline of DIN 13-21 international standard. Our procedure can potentially be used to restore and repair the dismantled spent fuel assembly.

The process basket assembly is an important module in pyroprocess, because pyroprocess is a batch process, so process materials are contained in a basket assembly and transferred with the basket. The basket assembly is composed of upper and lower assembly. The lower assembly is a basket or crucible which contains process materials, and it can have electrodes. The upper assembly mainly consists of heat shields, a flange, and connectors for supplying currents to electrodes of the lower assembly. During the electrolytic recovery process, the lower part is submerged into molten salt, whose temperature is about 500°C at least and the heat from salt is transferred to the upper assembly. And the heat affects the performance or durability of parts on the top of equipment and can raise cell temperature, which is an undesired situation. In addition, the handling equipment can pick the assembly when it is hottest, and during the transfer, the gripped part is under thermal and mechanical stresses. Because of this, the thermal effects from the heat should be required during equipment design stage. In this study, the thermal analyses of process basket assembly were conducted for 3 cases: the steady state of the basket assembly when it submerged in molten salt, the thermal and mechanical stresses when gripped by remote handling device, and the temperature changes under natural convection. These analyses were performed using Solidworks with flow simulation package, and the results will apply to improve the thermal resistant performance of the basket assembly.

Helicoverpa assulta (Lepidoptera: Noctuidae) exhibits a specialized herbivorous diet, primarily targeting select Solanaceae plants. Despite its significant economic impact as a pest, causing substantial harm to crops like hot pepper and tobacco, it has received comparatively limited attention in research compared to its generalist counterparts, H. armigera and H. zea.We introduce a chromosome level genome assembly using a Korean H. assulta (Pyeongchang strain, K18). This assembly was achieved through a combined approach utilizing Nanopore long-read sequencing (approximately 78X coverage) and Illumina NovaSeq short-read sequencing (approximately 54X coverage). The total assembled genome spans 424.36 Mb, designated as ASM2961881v1, comprises 62 scaffolds, with 98.7% of the genome contained within 31 scaffolds, confirming the insect's chromosome count (n = 31). The completeness of the assembly is reflected in BUSCO assessment, with values reaching 99.0%, while the repeat content accounts for 33.01%, and 18,593 CDS were annotated. Additionally, 137 genes were identified within 15 orthogroups that have rapidly expanded in H. assulta, while 149 genes in 95 orthogroups have rapidly contracted. This genome draft serves as a valuable resource to explore various aspects of the specialist's biology, enabling research into host-range evolution, chemical communication, insecticide resistance, and comparative investigations with other Heliothine species.

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO2 fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

An important goal of dismantling process is the disassembling of a spent nuclear fuel assembly for the subsequent extraction process. In order to design the rod extractor and cutter, the major requirements were considered, and the modularization design was carried out considering remote operation and maintenance. In order to design the rod extractor and cutter, these systems were analyzed and designed, also the concept on the rod extraction and cutting were considered by using the solid works tool. The main module consists of five sub-modules, and the function of each is as follows. The clamping module is an assembly fixing module using a cylinder so that the nuclear fuel assembly can be fixed after being placed. The Pusher module pushes the fuel rods by 2 inches out of the assembly to grip the fuel rods. The extraction module extracts the fuel rods of the nuclear fuel assembly and moves them to the consolidation module. The consolidation module collects and consolidates the extracted fuel rods before moving them to the cutting device. And the support module is a base platform on which the modules of the main device can be placed. The modules of level 2 can be disassembled or assembled freely without mutual interference. For the design of fuel rods cutter, the following main requirements were considered. The fuel rod cut section should not be deformed for subsequent processing, and the horizontally mounted fuel rods must be cut at regular intervals. The cutter should have the provision for aligning with the fuel rod, and the feeder and transport clamp should be designed to transfer the fuel rods to the cutting area. The main module consists of 6 sub-modules, and function of each is as follows. The cutting module is a device that cuts the fuel rods to the appropriate depth for notching. The impacting module is a device that impacts the fuel rods and moves them to the collection module. The transfer module is a device that moves the fuel rods to the cutting module when the aligned fuel rods enter the clamp module. The clamping module is a device to clamp the fuel rods before moving them to the cutting module. The collection module is a container where the rod-cuts are collected, and the support module is a base platform on which the modules of the main device can be placed. The module of level 3 can be disassembled or assembled after the cutting module of level 2 is installed, and the modules of level 2 can be disassembled or assembled freely without mutual interference.

This study investigates the behavior of the thermal conductivity among material properties in order to develop a thermal evaluation methodology of spent fuel assembles in a transport cask. It is inefficient to model each element of the spent fuel assembly in detail, and it is generally calculated by modeling the effective thermal conductivity (ETC). The ETC model was developed to allow a much simpler representation of a spent fuel assembly within a fuel compartment by treating the entire spent fuel rod array and the surrounding fill gas within the confines of the compartment as a homogenous solid material. The fuel rod assembly and surrounding gas are modeled with an effective conductivity that is designed to yield an overall conduction heat transfer rate that is equivalent to the combined effect of local conduction and radiation heat transfer in a plane through the assembly. When this model is applied to the transport cask, it tends to predict the cladding peak temperature lower than the results of detailed model in which the fuel rod arrangement and shape of the fuel assembly are simulated. As for the tendency of the error, the model tended to under-predict when basket temperature was lower than a certain temperature, and over-predict when it was higher. The purpose of this study is to investigate the attenuation effect of the cladding peak temperature on the related variables when the ETC model is applied to the transport cask. In addition, based on the thermal characteristics of this model, a correction factor that can compensate for this attenuation effect is presented. This correction factor is obtained by finding the difference between a separate ETC homogeneous model and a separate detailed fuel model, rather than directly applying the ETC calculated from the detailed fuel model to the transport cask.

Spent nuclear fuels released from the reactor are stored in cooling pools and then stored in dry storage casks. During the transition from the wet storage to dry storage cask, a vacuum drying process is used to remove residual water in the cask. During the vacuum drying process, gas pressure is reduced to below 400 Pa to promote evaporation and water removal. KAERI is developing a PWR single assembly (PLUS7) test equipment to simulate the thermal flow in spent fuel assembly. In this study, the thermal conductivity of air at low pressure was derived to perform the thermal analysis of the canister in vacuum. In addition, thermal analyses were performed for the canister with backfill gases of helium, air, and a vacuum in the vertical orientation using the COBRA-SFS code. At low pressure, the thermal conductivity of air depends on pressure and temperature. The reduced thermal conductivity, kr (W/m-K) was calculated using the curve fit for air at reduced pressure in thin gaps presented in the General Electric Fluid Flow Handbook. 􀝇􀯥/􀝇􀬴 = 􀬵 􀬵􀬾 􀮼􀯍/􀯉􀰋 Where, k0 is the thermal conductivity at atmospheric pressure (W/m-K), P is the reduced (vacuum) pressure (Pa), δ is the gap size (m), T is the temperature (K), and C is the Lasance constant (7.657E-5 N/m-K). The thermal conductivity of air decreases as the pressure decreases. The reduced thermal conductivity of air at pressures of 400 Pa and 40 Pa was calculated to be 0.97 and 0.77, respectively. For the analysis in vacuum, no enhancement of the convective heat transfer was assumed (Nu=1.0). For the helium backfill, the peak cladding temperature was the lowest and the axial temperature profile was the flattest due to the higher thermal conductivity and lower density of the helium. For the vacuum backfill, the peak cladding temperature was the highest and temperature gradient was the sharpest due to the only radiative heat transfer effect in the fuel assembly.