• Journal of the Korea Concrete Institute
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• v.22 no.3
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• pp.399-407
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• 2010

The flexural behaviors of two types of beam members exposed to freeze-thaw cycles were evaluated. This study aims to examine the effect of freeze-thaw cycles on the behavior characteristics of reinforced concrete (RC) beams. For the purpose, a part of the beam specimens were damaged until yielding of tension reinforcement was reached, before they were exposed to 150 and 300 cycles of freeze-thaw. Cyclic tests, as well as monotonic tests, were conducted to evaluate the stiffness degradation characteristics when same cycle is repeated. The material tests showed that relative dynamic modulus of concrete exposed to 300 cycles of freeze-thaw moderately decreased to 86.8% of normal concrete, indicating that concrete used in this study has good durability against freeze and thaw damage. The results of monotonic tests showed reduction of flexural strength, ductility and stiffness of the beam specimens exposed to freeze-thaw cycles compared with those of the control speciments. In particular, BDF13 specimens, which had been subjected to artificial cracking damage, did not showed enough flexural strength to satisfy nominal moment required by current concrete structure design code. In the monotonic tests results, BF75 specimens exposed to freeze-thaw cycles showed 10% or more cyclic stiffness degradation. Therefore, it was thought that deformation of concrete in compression have to be considered in design process of members under cyclic load, such as seismic device.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.6
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• pp.618-629
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• 2001

This paper presents a result of failure analysis and reliability evaluation for high voltage ceramic capacitors. The failure modes and failure mechanisms were studied in two ways in order to estimate component life and failure rate. The causes of failure mechanisms for zero resistance phenomena under withstanding voltage test in high voltage ceramic capacitors molded by epoxy resin were studied by establishing an effective root cause failure analysis. Particular emphasis was placed on breakdown phenomena at the ceramic-epoxy interface. The validity of the results in this study was confirmed by the results of accelerated testing. Thermal cycling test for high voltage ceramic capacitor mounted on a magnetron were implemented. Delamination between ceramic and epoxy, which might cause electrical short in underlying circuitry, can occur during curing or thermal cycle. The results can be conveniently used to quickly identify defective lots, determine $B_{10}$ life estimation each lot at the level of inspection, and detect major changes in the vendors processes. Also, the condition for dielectric breakdown was investigated for the estimation of failure rate with load-strength interference model.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.11
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• pp.144-150
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• 2019

In this study, resistance spot welding was performed using a high tensile steel plate SGAFC 780. The shear tensile strength, fracture profile, nugget diameter, and simulation were compared according to the conditions. After the nugget diameter calibration, the minimum diameter of welding was more than 4.3mm when the welding current was 8kVA or more. At 9kVA and above 10kVA, the minimum nugget diameter of 4.3mm was satisfied. On the other hand, due to the high current and time, the fly phenomenon occurred and the deep indentation remained. An evaluation of the weldability confirmed that there was an interval that was evaluated as weld failure due to the creep phenomenon, which satisfied the tensile shear strength and minimum nugget diameter. On the other hand, areas that have sufficient load bearing capacity even when drift has occurred were also identified. The simulation results show that the error rate was less than 4.2% when comparing the nugget diameter in the simulation and the experimental results in the appropriate weld zone, and confirmed the reliability of the simulation.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.21 no.11
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• pp.851-860
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• 2020

With an increase in the aging population and a rising social interest in health and welfare, studies to improve healthcare in the elderly are being actively conducted. This study attempted to improve the current design and manufacture of elevating electric wheelchairs to enhance user safety and convenience. Seat design based on the user's body shape, convenience while boarding or alighting, caster turning radius and, safety and stability features that prevent shaking when the user gets up or sits down were improved. A driving experiment was conducted to evaluate the operation of the indoor electric wheelchair designed and manufactured with these additional functionalities. During the test, the performance parameters evaluated were continuous driving time, turning radius, maximum lifting and lowering load, maximum lifting height, noise level, minimum distance sensing by the driving auxiliary sensor, ability to interact with server and app programs, and the duty cycle maximum error rate. The test confirmed that this improved electric wheelchair successfully met target parameters. In a future study, we will evaluate this improved electric wheelchair from a user's perspective for its usability parameters, such as satisfaction, convenience and stability.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.17-24
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• 2006

A demonstration facility for advanced spent fuel conditioning process (ACP) is under construction in KAERI. In this hot cell facility, all process equipments and materials are taken in and out only through the rear door. The working table in front of the process rear door is specially designed to be openable for the efficient use of the space. This paper presents the structural safety analysis of the openable working table, for the normal operational load condition and accidential drop condition of heavy object. Both cases are investigated through static and dynamic finite element analyses. The analysis results show that structural safety of the working table is sufficiently assured and the working table is not collapsed even when an object of 500 kg is dropped from the height of 50 cm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.229-238
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• 2007

In this paper, structural design requirements and safety evaluation criteria of the spent nuclear fuel disposal canister are studied for deep geological deposition. Since the spent nuclear fuel disposal canister emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal canister should be secured. Usually this repository is expected to locate at a depth of 500m underground. The canister which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock is a solid structure with cast iron insert, corrosion resistant overpack and lid and bottom, and entails an evenly distributed load of hydrostatic pressure from underground water and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. If the canister is not designed for all possible external loads combinations, structural defects such as plastic deformations, cracks, and buckling etc. may occur in the canister during depositing it in the deep repository. Therefore, various structural analyses must be performed to predict these structural problems like plastic deformations, cracks, and buckling. Structural safety evaluation criteria of the canister are studied and defined for the validity of the canister design prior to the structural analysis of the canister. And structural design requirements(variables) which affect the structural safety evaluation criteria should be discussed and defined clearly. Hence this paper presents the structural design requirements(variables) and safety evaluation criteria of the spent nuclear fuel disposal canister.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.301-309
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• 2006

KDC-1 canister for PWR spent fuels which will be used for the Korean Reference Disposal System was developed. The structural analysis of the canister was carried out as a part of the safety analysis. Two conditions, disposal condition and handling condition, were considered for the structural analysis. Three kinds of load cases, normal, abnormal and rock movement, were considered for the disposal condition. The results of the calculation showed that the safety factors from the structural analysis were greater than the design requirements. Two accident scenarios, gripper failure accident and canister drop accident, were analyzed for the handling condition. According to the gripper failure scenario analysis, the handling machine with grippers could be used even in the cases that one or two grippers failed. The maximum von Mises stress from the canister drop accident scenario was 0.762 MPa, which was negligible compared with the yield stress of nodular cast iron. The proposed KDC-1 canister for PWR spent fuels proves to be safe under the repository condition that is based upon the Korean reference disposal system according to the structural analysis for disposal condition and handling condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.37-43
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• 2012

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.229-236
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• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.