• Korean Membrane Journal
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• v.5 no.1
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• pp.68-74
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• 2003

Methane steam reforming (MSR) reaction for hydrogen production was studied in a membrane reactor (MR) using two tubular membranes, one Pd-based and one of porous alumina. A higher methane conversion than the thermodynamic equilibrium for a traditional reactor (TR) was achieved using MRs. The experimental temperature range was 350-500$^{\circ}C$; no sweep-gas was employed during reaction tests to avoid its back-permeation through the membrane and the steam/methane molar feed ratio (m) varied in the range 3.5-5.9. The best results (the difference between the MR conversion and the thermodynamic equilibrium was of about 7%) were achieved with the alumina membrane, working with the highest steam/methane ratio and at 450$^{\circ}C$. Silica membranes prepared at KRICT laboratories were characterized with permeation tests on single gases (N$_2$, H$_2$ and CH$_4$). These membranes are suited for H$_2$ separation at high temperature.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.237-245
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• 2018

Potential air ingress scenarios during accidents in nuclear reactors or spent fuel pools have raised the question of the influence of air, especially of nitrogen, on the oxidation of zirconium alloys, which are used as fuel cladding tubes and other structure materials. In this context, the reaction of zirconium with nitrogen-containing atmospheres and the formation of zirconium nitride play an important role in understanding the oxidation mechanism. This article presents the results of analysis of the interaction of the oxygen-preloaded niobium-bearing alloy $M5^{(R)}$ with nitrogen over a wide range of temperatures ($800-1400^{\circ}C$) and oxygen contents in the metal alloy (1-7 wt.%). A strongly increasing nitriding rate with rising oxygen content in the metal was found. The highest reaction rates were measured for the saturated ${\alpha}-Zr(O)$, as it exists at the metal-oxide interface, at $1300^{\circ}C$. The temperature maximum of the reaction rate was approximately 100 K higher than for Zircaloy-4, already investigated in a previous study. The article presents results of thermogravimetric experiments as well as posttest examinations by optical microscopy, scanning electron microscopy (SEM), and microprobe elemental analyses. Furthermore, a comparison with results obtained with Zircaloy-4 will be made.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.709-731
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• 2022

Loss of fracture resistance due to thermal aging degradation is a potential limiting factor affecting the long-term (80+ year) viability of nuclear reactors. To evaluate the effects of decades of aging in a practical time frame, accelerated aging must be employed prior to mechanical characterization. In this study, a variety of chemically and microstructurally diverse austenitic stainless steels were aged between 0 and 30,000 h at 290-400 ℃ to simulate 0-80+ years of operation. Over 600 static fracture tests were carried out between room temperature and 400 ℃. The results presented include selected J-R curves of each material as well as K_0.2mm fracture toughness values mapped against aging condition and ferrite content in order to display any trends related to those variables. Results regarding differences in processing, optimal ferrite content under light aging, and the relationship between test temperature and Mo content were observed. Overall, it was found that both the ferrite volume fraction and molybdenum content had significant effects on thermal degradation susceptibility. It was determined that materials with >25 vol% ferrite are unlikely to be viable for 80 years, particularly if they have high Mo contents (>2 wt%), while materials less than 15 vol% ferrite are viable regardless of Mo content.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1571-1578
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• 2020

As a critical material in very/high-temperature gas-cooled reactors, graphite material directly affects the safety of the reactor core structures. Owing to the complex structures of graphite material in reactors, the material typically undergoes complex stress states. It is, therefore, necessary to study its mechanical properties, failure modes, and strength criteria under complex stress states so as to provide guidance for the core structure design. In this study, compressive failure tests were performed for graphite material under the condition of different confining pressures, and the effects of confining pressure on the triaxial compressive strength and Young's modulus of graphite material were studied. More specifically, graphite material based on the fracture surfaces and fracture angles, the graphite specimens were found to exhibit four types of failure modes, i.e., tension failure, shear-tension failure, tension-shear failure and shear failure, with increasing confining pressure. In addition, the Mohr strength envelope of the graphite material was obtained, and different strength criteria were compared. It showed that the parabolic Mohr-Coulomb criterion is more suitable for the strength evaluation for the graphite material.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.11
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• pp.1125-1130
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• 2017

The tooth shapes of serration-type and spline-type reactors are optimized using finite element methods to improve the working life of the part and to lower the stress concentration during rotation resulting from contact with the outer race for a reactor operating with $170^{\circ}C$ transmission oil. The results of thermal expansion analyses between an Al reactor and the steel outer race indicate that, before optimization, the gap between the two parts increases further as the serration-type reactor expands by 0.1 mm and the spline-type one strains by 0.08 mm. Because of shape optimization, a trapezoidal shape is obtained from the initial triangular serration and the rectangular spline of the two reactors. The maximum von Mises stress of the serration-type convertor decreased by 24.5 ％, and by 9.3 ％ for the spline-type convertor. In addition, there is a 13 ％ reduction in the axial thickness, as compared to the initially designed model.

• Journal of Korean Society of Water and Wastewater
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• v.27 no.1
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• pp.21-29
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• 2013

Present study was conducted to evaluate the performance of Anaerobic Hybrid Reactor (AHR) combined with two types of anaerobic attached growth reactors at mesophilic temperature ($37^{\circ}C$). The reactor was operated at the influent substrate condition of 19,400 mg/L soluble chemical oxygen demand (sCOD). The organic loading rate (OLR) and flow rate were varied in the range of $9.5{\sim}22.5kg/m^3$. day and 10.6 ~ 26.0 L/day respectively since start-up was done. The COD removal efficiency of 93 % was measured at the OLR of $14kg/m^3$. day in AHR. However a reduction in removal efficiency to as low as 85 % could have been related to a combined effect of high concentration suspended solids (SS) concentration over 3,800 mg/L. On the other hand the total COD removal efficiencies were measured to be 96.3 % and 96.2 % for AHR+APF and AHR+ADF respectively. The pH of the POME was adjusted to neutral range by using sodium bicarbonate at the initial stages of the reactor feed, later stages pH adjustment was not required as the pH was maintained in the desired neutral range due to self-buffering capacity of the reactor. The reactor proved to be economically acceptable and operationally stable. The biogas was measured to have $CH_4$ and $CO_2$ with a ratio of 35:65, and methane gas production rate was estimated to be $0.17{\sim}10.269L\;CH_4/g\;COD_{removed}$.

• Journal of Environmental Health Sciences
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• v.32 no.4 s.91
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• pp.353-358
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• 2006

This study was performed to evaluate the effects of input ratios of bulking material in aerobic composting of food wastes on variation of colony forming units of heterotrophic bacteria. Wood chips were used as a bulking material. Volume ratios of food wastes to wood chips in reactor of Control, WC-1 and WC-2 were 10/0, 10/5 and 10/10, respectively. Reactors were operated for 24 days with 1hour stirring by 1rpm and 2 hours of the forced aeration per day. WC-2 reached high temperature range faster than WC-1, and the maximum temperature of WC-2 was higher than that of WC-1. This means that the reaction velocity of composting of WC-2 was faster than that of WC-1. Judging from the profile of pH changes, composting of WC-1 proceeded slowly and continuously. Composting of WC-2 proceeded rapidly in the former half reaction period, and kept steady state of high pH in the latter half reaction period. Namely, composting of WC-2 was nearly completed in the former half reaction period. In the case of WC-1 and WC-2. the maximum temperature was followed by the rapid pH increase in 2-3 days, and this was followed by the maximum Colony Forming Units(CFU) in 3 days. But, these three items of WC-2 always appeared faster and higher than those of WC-1.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.453-458
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• 2014

SiC and its composites have been considered for use as nuclear fuel cladding materials of pressurized light water reactors. In this study, a $SiC_f$/SiC composite as a constituent layer of SiC triplex fuel cladding was fabricated using a chemical vapor infiltration (CVI) process in which tubular SiC fiber preforms were prepared using a filament winding method. To enhance the matrix density of the composite layer, winding patterns, deposition temperature, and gas input ratio were controlled. Fiber arrangement and porosity were the main parameters influencing densification behaviors. Final density of the composites decreased as the SiC fiber volume fraction increased. The CVI process was optimized to densify the tubular preforms with high fiber volume fraction at a high $H_2$/MTS ratio of 20 at $1000^{\circ}C$; in this process, surface canning of the composites was effectively retarded.

• Journal of the Korean Ceramic Society
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• v.43 no.7 s.290
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• pp.439-444
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• 2006

Artificial graphite generally manufactured by carbonization sintering of shape-body of kneaded mixture using granular cokes as filler and pitch as binder, going through pitch impregnation process if necessary and finally applying graphitization heat treatment. Graphite materials are used for core internal structural components of the High-Temperature Gas-cooled Reactors (HTGR) because of their excellent heat resistibility and resistance of crack progress. The HTGR has a core consisting of an array of stacked graphite fuel blocks are machined from IG-110, a high-strength, fine-grained isotropic graphite. In this study, crack stabilization and micro-structures were measured by bend strength and fracture toughness of isotropic graphite grade IG-110. It is important to the reactor designer as they may govern the life of the graphite components and hence the life of the reactor. It was resulted crack propagation, bend strength, compressive strength and micro-structures of IG-110 graphite by scanning electron microscope and universal test machine.