• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.83-88
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• 2011

To define the effect of strain rate variation from 0.04% to 0.004%/s on environmental fatigue of CF8M cast stainless steel, which is used as a primary piping material in nuclear power plants, low-cycle fatigue tests were conducted at operating pressure and temperature condition of a pressurized water reactor, 15 MPa and $315^{\circ}C$, respectively. A high-pressure and high-temperature autoclave and cylindrical solid fatigue specimens were used for the strain-controlled low-cycle environmental fatigue tests. It was observed that the fatigue life of CF8M stainless steel is shortened as the strain rate decreases. Due to the effect of test temperature, the fatigue data of NUREG-6909 appears a slightly shorter than that obtained by KEPRI at the same stress amplitude of $1{\times}10^3$ MPa. The environmental fatigue correction factor $F_{en}$'s calculated with inputs of the test data increases with high strain amplitude, while the $F_{en}$'s of NUREG-6909 remain constant regardless of strain amplitude.

• Journal of the Korean Solar Energy Society
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• v.27 no.1
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• pp.75-81
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• 2007

Steam reforming of methane using Xe-arc solar simulator was studied for the application of concentrated solar energy into chemical reaction. The reactor, a volumetric absorber, consisted of a porous ceramic foam disk coated with commercial reforming catalyst. Operating temperature was in the range of $450\;-\;550^{\circ}C$ and the excess steam ratio to methane was from 3.0 to 5.0. At the steady-state condition, the conversion of methane Increased with temperature in the range of 15 % - 30 % and the experimentally determined conversion was found to be close to theoretical equilibrium conversion. It was also found that the CO selectivity slightly decreased with excess steam ratio. Finally, the conversion of methane decreased significantly with space velocity of reactants.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1596-1609
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• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.47-54
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• 2013

This study investigates the crack growth behavior due to primary water stress corrosion cracking (PWSCC) in the dissimilar metal butt weld of a reactor piping using Alloy 82/182. First, detailed finite element stress analyses were performed to predict the stress distribution of the dissimilar metal butt weld in which the hydrostatic and the normal operating loads as well as the weld residual stresses were considered to evaluate the stress redistribution due to mechanical loadings. Based on the stress distributions along the wall thickness of the dissimilar metal butt weld, the crack growth behavior of the postulated axial and circumferential cracks were predicted, from which the crack growth diagram due to PWSCC was proposed. The present results can be applied to predict the crack growth rate in the dissimilar metal butt weld of reactor piping due to PWSCC.

• Applied Chemistry for Engineering
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• v.23 no.1
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• pp.28-34
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• 2012

In this study, we operated a UASB (upflow anaerobic sludge blanket) reactor by using foodwaste leachate as a raw material with the method of Mesophilic Digestion ($35{\pm}0.5^{\circ}C$) and Thermophilic Digestion ($55{\pm}0.5^{\circ}C$). During 20 days of operating time with the Mesophilic Digestion, the recirculation ratio of effluent was stepwisely changed in every five days. Thermophilic Digestion was carried out at the same condition for Mesophilic Digestion. Results showed that the organic removal efficiency of Mesophilic Digestion was over 90% and the yield of methane production was from 66 up to 70%. The organic removal efficiency of Thermophilic Digestion was over 80% and the yield of methane production was between 62 to 68%. Also, when UASB reactor was operating to over the 3Q effluent recirculation, the experiment could be carried out economically and stably.

• Journal of the Korea Organic Resources Recycling Association
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• v.15 no.3
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• pp.121-128
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• 2007

The purpose of this study was to investigate the removal efficiencies of HCl in dry reactor and Bag-Filter system. The bicarbonate was used as adsorbent to measure the HCl removal rate. The performance of bicarbonate was evaluated to investigate the removal efficiency. It was analysed that the best operation condition in using bicarbonate in process. The operating parameters was residence time, stoichiometric ratio, temperature and pressure. The Residence time was 1.5 sec, stoichiometric ratio was 1SR, 1.25SR, temperature was $160^{\circ}C$, $180^{\circ}C$, $200^{\circ}C$ and the bag filter pressure was 210mmAq, 230mmAq, 250mmAq. In this study it was found that the stoichiometric ratio and pressure can be effected in removal of HCl however the removal efficiency do not depends on temperature. The best operating condition was on 1SR, 230mmAq and $180^{\circ}C$ respectively and in this condition the best removal rate was 99.50%.

• Korean Chemical Engineering Research
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• v.44 no.5
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• pp.528-534
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• 2006

The purpose of this paper is to investigate the optimal operating condition for the hydrogen production by biogas reforming using the plasmatron induced thermal plasma. The component ratio of biogas($CH_4/CO_2$) produced by anaerobic digestion reactor were 1.03, 1.28, 2.12, respectively. And the reforming experiment was performed. To improve hydrogen production and methane conversion rates, parametric screening studies were conducted, in which there are the variations of biogas flow ratio(biogas/TFR: total flow rate), vapor flow ratio($H_2O/TFR$: total flow rate) and input power. When the variations of biogas flow ratio, vapor flow ratio and input power were 0.32~0.37, 0.36~0.42, and 8 kW, respectively, the methance conversion reached its optimal operating condition, or 81.3~89.6%. Under the condition mentioned above, the wet basis concentrations of the synthetic gas were H2 27.11~40.23%, CO 14.31~18.61%. The hydrogen yield and the conversion rate of energy were 40.6~61%, 30.5~54.4%, respectively, the ratio of hydrogen to carbon monoxide($H_2/CO$) was 1.89~2.16.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.343-354
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• 2012

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.