• Korean Journal of Metals and Materials
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• v.47 no.3
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• pp.139-146
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• 2009

The heat-affected zone (HAZ) of SA508 Gr.4N Ni-Mo-Cr low alloy steel, which has higher Ni and Cr contents than SA508 Gr.3 Mn-Mo-Ni low alloy steel, was investigated on the microstructure and mechanical properties. The HAZ was categorized into seven characteristic zones (CGCG, FGCG, ICCG, SCCG, FGFG, ICIC and SCSC-HAZ) according to the peak temperature from the thermal cycle experienced during multi-pass welding. Post Weld Heat Treatment (PWHT) was conducted in the temperature range of $550{\sim}610^{\circ}C$ for 30 hours to evaluate the effect of PWHT conditions on the microstructure and mechanical properties. Before PWHT, CGHAZ and FGFGHAZ showed high yield strength (YS) ranging from 1000 to 1250 MPa, while YS of SCSCHAZ decreased from 607 MPa (observed for base metal) to 501 MPa. The Charpy impact energies of sub-HAZs fell below 100J at $-29^{\circ}C$, except in the SCSCHAZ. By applying PWHT to sub-HAZ specimens, YS decreased as the PWHT temperature increased. In the case of CGHAZs and FGFGHAZ heat-treated at $610^{\circ}C$, YS dropped drastically to the range of 654~686 MPa. From the Charpy impact test, the upper-shelf energy (USE) increased to approximately 250J and Index temperature ($T_{68J}$) decreased below $-50^{\circ}C$. Specifically, in FGFG, ICIC and SCSC-HAZ, $T_{68J}$ was below -110, which was lower than the case of base metal.

• Journal of the Korean Institute of Gas
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• v.25 no.1
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• pp.40-47
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• 2021

In this study, hazards of decomposition and explosion for tert-butylperoxymaleate(TBPM), an organic peroxide, were evaluated by using various equipment to determine the cause of a fire explosion accident. As a result of DSC analysis, the instantaneous power density of TBPM was 26,401 kW/ml, and the NFPA reactive index(Nr) was classified as 4. And the positive value of EP(explosive propagation) and SS(shock sensitivity) showed that the TBPM had a potential hazard of explosion. From the experimental results, the shock sensitivity and friction sensitivity was rated as class 4 and 5, respectively. In the pressure vessel test, TBPM was ranked USA-PVT No.4 and evaluated as a self-reactive substance. In the combustion rate test, TBPM had the combustion rate of 167 mm/sec and was evaluated as the flammable solid classification 2 in GHS.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1651-1664
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• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

• Journal of Korean Neurosurgical Society
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• v.67 no.2
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• pp.227-236
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• 2024

Objective : Numerous studies have indicated that early decompressive craniectomy (DC) for patients with major infarction can be life-saving and enhance neurological outcomes. However, most of these studies were conducted by neurologists before the advent of intra-arterial thrombectomy (IA-Tx). This study aims to determine whether neurological status significantly impacts the final clinical outcome of patients who underwent DC following IA-Tx in major infarction. Methods : This analysis included 67 patients with major anterior circulation major infarction who underwent DC after IA-Tx, with or without intravenous tissue plasminogen activator. We retrospectively reviewed the medical records, radiological findings, and compared the neurological outcomes based on the "surgical time window" and neurological status at the time of surgery. Results : For patients treated with DC following IA-Tx, a Glasgow coma scale (GCS) score of 7 was the lowest score correlated with a favorable outcome (p=0.013). Favorable outcomes were significantly associated with successful recanalization after IA-Tx (p=0.001) and perfusion/diffusion (P/D)-mismatch evident on magnetic resonance imaging performed immediately prior to IA-Tx (p=0.007). However, the surgical time window (within 36 hours, p=0.389; within 48 hours, p=0.283) did not correlate with neurological outcomes. Conclusion : To date, early DC surgery after major infarction is crucial for patient outcomes. However, this study suggests that the indication for DC following IA-Tx should include neurological status (GCS ≤7), as some patients treated with early DC without considering the neurological status may undergo unnecessary surgery. Recanalization of the occluded vessel and P/D-mismatch are important for long-term neurological outcomes.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• Journal of the Society of Naval Architects of Korea
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• v.61 no.3
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• pp.161-169
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• 2024

In order to investigate cavitation performance for the azimuth thruster in Large Cavitation Tunnel (LCT), the cavitation test apparatus was designed and manufactured. Generally the model scale is determined by the pod dynamometer with about 70mm diameter. Recently as ships with azimuth thruster have become bigger, the problem of the model ship installation was occurred. The model ship with pod dynamometer couldn't be installed in the LCT test section. The cavitation test apparatus and technique which can conduct the cavitation test without pod dynamometer were developed. The cavitation tests were conducted in torque identity method instead of thrust identity method. The target ship with azimuth thruster is 18K LNG bunkering vessel. As the full-scale ship test was conducted, the model cavitation tests were conducted at the same operating conditions. The fluctuating pressure levels of the full-scale ship were compared to those of the model tests. Another model cavitation test was conducted in the foreign institute and the cavitation observation results were compared to those of LCT. Through the comparison with the existing results, it is thought that the cavitation test for the azimuth thruster can be conducted in torque identity method.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• Journal of Korean Biological Nursing Science
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• v.4 no.1
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• pp.71-83
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• 2002

The purpose of this study was to determine the effect of exercise designed on physiological variables, physical functions in the elderly. The research design used was the one group pre-post test quasi-experimental design. The subjects for this study were 23 elderly, who were conveniently assigned to experimental group. The experimental group were selected from G-elderly church club in G-city. The data were collected from March 5, 2001 to June 11, 2002, and analyzed by frequency, t-test using SPSS/PC program. The results of this study were summarized as follows ; 1) The heart rate in the experimental group show statistically significant improvement (t=2.316, p=0.030). 2) The systolic blood pressure & diastolic blood pressure in the experimental group show statistically significant improvement(t=2.277, p=0.033 & t=-2.989, p=0.007). 3) The peripheral vessel $O_2$ Saturation in the experimental group don't show statistically significant improvement(t=-1.096, p=0.285). 4) The physical function in the experimental group don't show statistically significant improvement(t=-0.746, p=0.464). After 15 weeks of exercise designed, it was shown that the program was effective to improve the physiological variables.

• Magazine of the Korean Society of Agricultural Engineers
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• v.39 no.4
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• pp.106-113
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• 1997

Dynamic loading of structures often causes excursions of stresses well into the inelastic range, and the influence of the geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. A structure in a nuclear power plant is a structure of importance which puts emphasis on safety. A nuclear container is a pressure vessel subject to internal pressure and this structure is constructed by a reinforced concrete or a pre-stressed concrete. In this study, the material nonlinearity effect on the dynamic response is formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion is numerically solved by a central difference scheme. The constitutive relation of concrete is modeled according to a Drucker-Prager yield criterion in compression. The reinforcing bars are modeled by a smeared layer at the location of reinforcements, and the steel layer model under Von Mises yield criteria is adopted to represent an elastic-plastic behavior. To investigate the dynamic response of a nuclear reinforced concrete containment structure, the steel-ratios of 0, 3, 5 and 10 percent, are considered. The results obtained from the analysis of an example were summarized as follows 1. As the steel-ratio increases, the amplitude and the period of the vertical displacements in apex of dome decreased. The Dynamic Magnification Factor(DMF) was some larger than that of the structure without steel. However, the regular trend was not found in the values of DMF. 2. The dynamic response of the vertical displacement and the radial displacement in the dome-wall junction were shown that the period of displacement in initial step decreased with the steel-ratio increases. Especially, the effect of the steel on the dynamic response of radial displacement disapeared almost. The values of DMF were 1.94, 2.5, 2.62 and 2.66, and the values increased with the steel-ratio. 3. The characteristics of the dynamic response of radial displacement in the mid-wall were similar to that of dome-wall junction. The values of DMF were 1.91, 2.11, 2.13 and 2.18, and the values increased with the steel-ratio. 4. The amplitude and the period of the hoop-stresses in the dome, the dome-wall junction, and the mid-wall were shown the decreased trend with the steel-ratio. The values of DMF were some larger than those of the structure without steel. However, the regular trend was not found in the values of DMF.