• Korean Chemical Engineering Research
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• v.59 no.1
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• pp.85-93
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• 2021

For accurate and reliable process design for phenol oxidation in a plug flow reactor with supercritical water, modeling can be very insightful. Here, the velocity and density distribution along the reactor have been predicted by a numerical model and variations of temperature and phenol mass fraction are calculated under various flow conditions. The numerical model shows that as we proceed along the length of the reactor the temperature falls from above 430 ℃ to approximately 380 ℃. This is because the generated heat from the exothermic reaction is less that the amount lost through the walls of the reactor. Also, along the length, the linear velocity falls to less than one-third of the initial value while the density more than doubles. This is due to the fall in temperature which results in higher density which in turn demands a lower velocity to satisfy the continuity equation. Having a higher oxygen concentration at the reactor inlet leads to much faster phenol destruction; this leads to lower capital costs (shorter reactor will be required); however, the operational expenditures will increase for supplying the needed oxygen. The phenol destruction depends heavily on the kinetic parameters and can be as high as 99.9%. Using different kinetic parameters is shown to significantly influence the predicted distributions inside the reactor and final phenol conversion. These results demonstrate the importance of selecting kinetic parameters carefully particularly when these predictions are used for reactor design.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.11a
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• pp.63-64
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• 2008

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.2
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• pp.191-200
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• 2011

A PHE (Process Heat Exchanger) is a key component of nuclear hydrogen system for massive production of hydrogen; the PHE transfers the very high temperature heat ($950^{\circ}C$) generated from the VHTR (Very High Temperature Reactor) to a chemical reaction. The Korea Atomic Energy Research Institute developed a small-scale gas loop for testing the performance of VHTR components and manufactured a modified PHE prototype for carrying out the testing in the gas loop. In this study, as a part of the evaluation of the high-temperature structural integrity of the modified PHE prototype which is scheduled to test in the gas loop, we carried out high-temperature structural analysis modeling, macroscopic thermal and structural analysis of the PHE prototype under the gas loop test conditions as a precedent study before carrying out the performance test in the gas loop. The results obtained in this study will be used to design the performance test setup for the modified PHE prototype.

• Journal of the Korean Ceramic Society
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• v.43 no.12 s.295
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• pp.833-838
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• 2006

Graphite is suitable for high temperature structural materials because of chemical stability as well as unique crystal structure. Especially, graphite can be used as a part of a nuclear reactor due to high tolerance at the extreme conditions of high temperature and neutron irradiations. Although study of oxidation properties or behaviors of graphite are very important and essential for the life and stability of the nuclear reactor, most of studies treat this theme lightly. This work focuses on the oxidation characteristics of several grade isotropic graphite of the nuclear reactor.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.566-584
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• 2001

In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53$0^{\circ}C$ and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.

• Korean Journal of Metals and Materials
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• v.50 no.9
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• pp.659-667
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• 2012

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at $850^{\circ}C-950^{\circ}C$ in a helium environment containing the impurity gases $H_2$, CO, and $CH_4$, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high-temperature corrosion behavior of Alloy 617 for the VHTR application.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.10
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• Journal of Ocean Engineering and Technology
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• v.18 no.2
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• pp.58-63
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• 2004

Reduced Activation Ferritic/Martensitic Steels (RAFs) are leading candidntes for structural materials of a D-T fusion reactor. One of the RAFs, JLF-l (9Cr-2W-V, Ta) has been developed and has shown to have good resistance against high-fluency neutrino irradiation and good phase stability. Recently, in order to clarify the strengthening mechanisms at high temperatures, a new scheme to improve high temperature mechanical properties is desired. Therefore, the test technique development of high temperature creep behaviors for this material is very important. In this paper, the creep properties and creep life prediction, using the Larson-Miler parameter method for JLF-l to be used for fusion reactor materials or other high temperature components, are presented at the elevated temperatures of 50$0^{\circ}C$, 55$0^{\circ}C$, $600^{\circ}C$, $650^{\circ}C$ and 704$^{\circ}C$. It was confirmed, experimentally and quantitatively, that a creep life predictive equation, at such various high temperatures, is well derived mr the LMP method.

• Journal of the Korean Ceramic Society
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• v.48 no.1
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• pp.52-56
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• 2011

Silicon carbide (SiC) is a candidate material for heat exchangers for VHTR (Very High Temperature Gas Cooled Reactor) due to its refractory nature and high thermal conductivity. This research has focused on demonstration of physical properties and mock-up fabrication for the future heat exchange applications. It was found that the SiC-based components can be applied for process heat exchanger (PHE) and intermediate heat exchanger (IHX), which are operated at $400{\sim}1000^{\circ}C$, based on our examination for the following aspects: optimum fabrication technologies (design, machining and bonding) for compact design, thermal conductivity, corrosion resistance in sulfuric acid environment at high temperature, and simulation results on heat transferring and thermal stress distribution of heat exchanger mock-up.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.71-79
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• 2023

Air ingress incidents are major safety accidents in very-high-temperature reactors (VHTRs). Air containing a high volume fraction of oxygen may cause severe oxidation of core components at the VHTR, especially for the significantly thin alloy tube wall in the intermediate heat exchanger (IHE). The research objects of this study are Inconel 617 and Incoloy 800H, two candidate alloys for IHE in VHTR. The air ingress accident scenario is simulated with high-temperature air flow at 950 ℃. A continuous oxide scale was formed on the surfaces of both the alloys after the experiment. Because the oxide scale of Inconel 617 has a loose structure, whereas that of Incoloy 800H is denser, Inconel 617 exhibited significantly more severe internal oxidation than Incoloy 800H. Further, Inconel 617 showed a significant decrease in ultimate tensile strength and plasticity after aging for 200 h, whereas Incoloy 800H maintained its tensile properties satisfactorily. Through control experiment under vacuum, we preliminarily concluded that serious internal oxidation is the primary reason for the decline in the tensile properties of Inconel 617.