• Proceedings of the KIEE Conference
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• 2001.07d
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• pp.2166-2168
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• 2001

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics. The Reactor internal structures which consist of many complex components are subjected to flow-induced vibration due to high temperature and pressure in reactor coolant system. The above flow-induced vibration causes degradation of structural integrity of the reactor and may result in loosing mechanical binding component which might impact other equipment and component or cause flow blockage. It is important to analyze reactor noise signal for the early detection of potential problem or failure in order to diagnosis reactor integrity in the point of view of safety and plant economics. Detailed design of hardware diagnostic system reactor internal structures using neutron noise(RIDS).

• 한국신재생에너지학회:학술대회논문집
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• 2008.05a
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• pp.210-213
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• 2008

The purpose of this study is to investigate the performances of organic removal and methane recovery in the full scale two-phase anaerobic system. The full scale two-phase anaerobic system was consists of an acidogenic ABR (Anaerobic Baffled Reactor) and a methanognic UASB (Upflow Anaerobic Sludge Blanket) reactor. The volume of acidogenic and methanogenic reactors is designed to 28.3 $m^3$ and 75.3 $m^3$. The two-phase anaerobic system represented 60-82% of COD removal efficiency when the influent COD concentration was in the range of 7,150 to 16,270 mg/L after screening (average concentration is 10,280 mg/L). After steady-state, the effluent COD concentration in the methanogenic reactor showed 2,740 $\pm$ 330 mg/L by representing average COD removal efficiency was 71.4 $\pm$ 8.1% when the operating temperature was in the range of 19-32$^{\circ}C$. The effluent SCOD concentration was in the range of 2,000-3,000 mg/L at the steady state while the volatile fatty concentration was not detected in the effluent. Meanwhile, the COD removal efficiency in the acidogenic reactor showed less than 5%. The acidogenic reactor played key roles to reduce a shock-loading when periodic shock loading was applied and to acidify influent organics. Due to the high concentration of alkalinity and high pH in the effluent of the methanogenic reactor, over 80% of methane in the biogas was produced consistently. More than 70 % of methane was recovered from theoretical methane production of TCOD removed in this research. The produced gas can be directly used as a heat source to increase the reactor temperature.

• Plant Journal
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• v.4 no.2
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• pp.60-65
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• 2008

As the operating conditions in a supercritical oxidation reactor are set in high temperature with high pressure causing a reactor suffering from the harsh circumstances. It means the reactor adopts itself with Fe-Cr alloy in acidic atmosphere with low pH value and Ni alloy in basic atmosphere with high pH value due to its superior corrosion resistance. The study, whose target waster water is pertinent to the latter part, has selected Ni alloy such as ostenite type stainless steel 304 and 316, superstainless steel AL6XN, Inconel 625, MAT 21, and titanium Gr. 5 in order to measure corrosion resistance against those samples under the same conditions of temperature and pressure applied for a supercritical oxidation reactor. The result shows the identifiable difference in corrosion resistance by observing the surface states through a scanning probe microscope as well as measuring the weight loss through making the samples above deposited in wastewater for two-week and four-week stay. The purpose of this corrosion experiment is to identify the most corrosion-resistant material among sample species pre-selected according to pH concentration of wastewater in pursue of applying for a reactor exposed to the extreme corrosion environment. It is because such a reactor made of a verified material enables to safeguard a stable operation under the supercritical wastewater processing facility.

• Journal of Mechanical Science and Technology
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• v.20 no.12
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• pp.2061-2078
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• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.1-6
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• 2011

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established a small-scale gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype made of Hastelloy-X to be tested in the small-scale gas loop. Results from the elastic structural analysis on the PHE prototype were reported in the previous article. In order to investigate the macroscopic structural characteristics and behavior of the PHE prototype under the test condition of the small-scale gas loop far more in detail, elastic-plastic high-temperature structural-analysis of the PHE prototype was carried out in this study.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.15-23
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• 2018

A compact heat exchanger is one of critical components in a very high temperature gas-cooled reactor (VHTR). Alloy 617 (Ni-Cr-Co-Mo) is considered as one of leading candidates for this application due to its excellent thermal stability and strengths in anticipated operating conditions. On the basis of current ASME code requirements, sixty sheets of this alloy are prepared for diffusion welding, which is the key technology to have a reliable compact heat exchanger. Optical microscopic analysis show that there are no cracks, incomplete bond, and porosity at/near the interface of diffusion weldment, but Cr-rich carbides and Al-rich oxides are identified through high resolution electron microscopic analysis. In high-temperature tensile testing, superior yield strengths of the diffusion weldment compared to the code requirement are obtained up to 1223 K ($950^{\circ}C$). However, both tensile strength and ductility drop rapidly at higher temperature due to the insufficient grain boundary migration across the interface of diffusion weldment. Best fit curves for minimum yield strength and average tensile strength are drawn from the experimental tensile results of this study.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.1
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• pp.33-38
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Transactions of the Korean hydrogen and new energy society
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• v.34 no.3
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• pp.267-273
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• 2023

As the impacts of global climate change become increasingly apparent, the reduction of carbon emissions has emerged as a critical subject of discussion. Nuclear power has garnered attention as a potential carbon-free energy source; however, the rapidity of load following in nuclear power generation poses challenges in comparison to fossil-fueled methods. Consequently, power-to-gas systems, which integrate nuclear power and hydrogen, have attracted growing interest. This study presents a preliminary design of a very high temperature reactor (VHTR) integrated blue hydrogen production process utilizing DWSIM, an open-source process simulator. The blue hydrogen production process is estimated to supply the necessary calorific value for carbon capture through tail gas combustion heat. Moreover, a thermodynamic assessment of the main recuperator is performed as a function of the helium flow rate from the VHTR system to the blue hydrogen production system.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.324-328
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• 2003

Water-lubricated frictional characteristics of stainless steel ball bearing is not well known compared to oil-lubricated frictional characteristics. Furthermore study on friction at high temperature is rare because bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is mostly based on change of failed bearings and parts. Ball bearings and ball screw are installed on the power transmission for a developing integral reactor and these are lubricated with high temperature and high pressure chemically-controlled pure water. Bearings and power transmitting mechanical elements for an atomic reactor needs high reliability. and high performance during estimated lifetime, and it should be verified. In this paper, experimental research results of frictional characteristics of water-lubricated ball bearing as a preliminary investigation.