• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2696-2707
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• 2021

The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer shell model (MLSM), and beam-truss model (BTM), are developed to simulate the seismic behaviors of the containment structure. A full three-dimensional finite element model (full 3D FEM) is additionally constructed to verify the previous numerical models. A set of input ground motions with response spectra matching to the US NRC 1.60 design spectrum is generated to perform linear and nonlinear time-history analyses. Floor response spectra (FRS) and floor displacements are obtained at the different elevations of the structure since they are critical outputs for evaluating the seismic vulnerability of RCB and secondary components. The results show that the difference in seismic responses between linear and nonlinear analyses gets larger as an earthquake intensity increases. It is observed that the linear analysis underestimates floor displacements while it overestimates floor accelerations. Moreover, a systematic assessment of the capability and efficiency of each structural model is presented thoroughly. MLSM can be an alternative approach to a full 3D FEM, which is complicated in modeling and extremely time-consuming in dynamic analyses. Specifically, BTM is recommended as the optimal model for evaluating the nonlinear seismic performance of NPP structures. Thereafter, linear and nonlinear BTM are employed in a series of time-history analyses to develop fragility curves of RCB for different damage states. It is shown that the linear analysis underestimates the probability of damage of RCB at a given earthquake intensity when compared to the nonlinear analysis. The nonlinear analysis approach is highly suggested for assessing the vulnerability of NPP structures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.95-106
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• 2019

The decommissioning of nuclear power plants should be prepared by quantitative and qualitative risk assessment. Radiological and non-radiological hazards arising during decommissioning activities must be assessed to ensure the safety of decommissioning workers and the public. Decommissioning experiences by U.S. operators have mainly focused on deterministic risk assessment, which is standardized by the U.S. Nuclear Regulatory commission (NRC) and focuses only on the consequences of risk. However, the International Atomic Energy Agency (IAEA) has suggested an alternative to the deterministic approach, called the risk matrix technique. The risk matrix technique considers both the consequence and likelihood of risk. In this study, decommissioning stages, processes, and activities are organized under a work breakdown structure. Potential accidents in the decommissioning process of NPPs are analyzed using the composite risk matrix to assess both radiological and non-radiological hazards. The levels of risk for all potential accidents considered by U.S. NPP operators who have performed decommissioning were estimated based on their consequences and likelihood of events.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.21 no.4
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• pp.125-136
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• 2011

The vulnerabilities existed in Korean electricity control systems is unexposed because it is being operated in a closed network with superior security. The threat will become greater once the closed network develops into a smart grid environment with superior intelligence. Security will have a greater impact once each household will be connected to the power plant via the smart meter. This research focuses on stable data transfer in harsh external environment and whole-nation coverage network, and suggested standardized and optimized Virtual Private Network (VPN) Gateway architecture to support Power Line Communication (PLC). The functionality and stability of the prototype has been verified with field tests. For implementation of outdoor type VPN device for smart grid, we adopted PLC low voltage remote-meter-net for data communication. Also, IPSec type tunneling and ARIA algorithm based encryption of data collected by PLC low voltage remote meter is transmitted.

• Journal of the Earthquake Engineering Society of Korea
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• v.26 no.1
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• pp.13-22
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• 2022

This study proposes a low-cycle fatigue life derived from measurement points on pipe elbows, which are components that are vulnerable to seismic load in the interface piping systems of nuclear power plants that use seismic isolation systems. In order to quantitatively define limit states regarding leakage, i.e., actual failure caused by low-cycle fatigue, in-plane cyclic loading tests were performed using a sine wave of constant amplitude. The test specimens consisted of SCH40 6-inch carbon steel pipe elbows and straight pipes, and an image processing method was used to measure the nonlinear behavior of the test specimens. The leakage lines caused by low-cycle fatigue and the low-cycle fatigue curves were compared and analyzed using the relationship between the relative deformation angles, which were measured based on each of the measurement points on the straight pipe, and the moment, which was measured at the center of the pipe elbow. Damage indices based on the combination of ductility and dissipation energy at each measurement point were used to quantitatively express the time at which leakage occurs due to through-wall cracking in the pipe elbow.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.43 no.6
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• pp.699-708
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• 2023

Nuclear power plants (NPPs) consider possible external events, including natural hazards, during the design phase to ensure safe operation. However, in recent years, due to the increasing probability of natural hazards exceeding the design, a careful review of extreme natural hazards and unforeseen external events during the design phase has become necessary. In this study, the objective was to screen potential extreme natural hazards at NPP sites in Korea. Initially, we investigated and analyzed the characteristics of NPP sites and the events caused by external hazards. Furthermore, we analyzed existing literature and research data to establish screening procedures and criteria that suit the actual conditions of domestic NPPs. Based on these criteria and data, we conducted qualitative screening for each NPP site and identified potential extreme natural hazards through quantitative screening and walkdown. As a result of the screening, in addition to internal flooding caused by heavy rain, wind pressure and extreme air pressure caused by extreme winds were screened as potential extreme natural hazards common to all sites. Additionally, at the Kori site, storm surge was selected as the most significant potential extreme natural hazard.

• Korean Journal of Soil Science and Fertilizer
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• v.25 no.2
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• pp.143-148
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• 1992

Fly ash treatment on soil had a strongly positive effect on the growth of soybean. Treatment of fly ash to the soil made soil pH improved and available phosphate content increased. Consequently yield of soybean increased. From germination to early growth stage, growth status and weight of the plant were unfavorably affected by fly ash and its effects on the leaf was quite serious specially in the plots treated with more than 10 MT/10a of bituminous fly ash. However after early stage, plant growth became vigorous in the order of 0 (control plot)<15<5<10 MT/10a. But at the late maturing stage, deteriorative symptoms such as leaf burn and drying were appeared from the plant treated with 10MT/10a and its symptoms were more serious with 15MT/10a. By anthracite fly ash treatment, the plant growth was greatly improved. As a result plant height and dry matter were in the order of 0<5<10<15MT/10a. Grain yield was in the order of 0<15<5< 10MT/10a treatment with bituminous fly ash and 0<5<10<15MT/10a treatment with anthracite fly ash. As a conclusion, recommandable amount of fly ash treatment for soybean would be 5-10 MT/10a with anthracite fly ash and 5 MT/10a with bituminous fly ash.

• International Journal of Air-Conditioning and Refrigeration
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• v.11 no.2
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• pp.51-60
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• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. Standard k-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• Journal of computational fluids engineering
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• v.16 no.2
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• pp.66-74
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• 2011

The SCR catalyst in coal-fired power plant is eroded by the collision of fly ash on the catalyst surface. However the erosion of SCR catalyst by the collision of fly ash has not been fully studied, especially in terms of fluid dynamics. Hence, in the present study, we focus on the gas and solid flows inside the SCR catalyst duct and their consequent effect on the erosion characteristics. For this purpose, computational fluid dynamics is applied to investigate the two-phase flows and to evaluate the erosion rate for different flow and particle injection conditions. Also, the erosion rate and pressure drop of commonly used square shape are compared with equilateral triangle and hexagon shapes. The pressure drop of SCR catalyst is increased when SCR catalyst surface area per unit volume increases. The erosion rate of SCR catalyst is enhanced when the particle velocity, mass flow rate of particle, particle diameter and cell density of SCR catalyst are increased. From the results, the pressure drop and erosion rate at the catalyst surface can be minimized by reducing cell density of SCR catalyst to decrease particle velocity and number of particle impacts.

• Tribology and Lubricants
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• v.25 no.5
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• pp.330-334
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• 2009

Hydraulic servo cylinders have been used to control accurately a large machine in power plant. Especially, Piston head and bush of servo cylinder is assembled sleeve and piston head and bush made of Cu alloy and pad sealing part. A damages of sleeve and piston head, bush are caused by friction and wear. Thus, It is necessary to examine friction and wear characteristics of Cu alloys for the piston head and bush. In this study, to be reliable on the piston and cylinder parts, dry friction and wear experiments were carried out with Cu alloys of four kinds of AlBC, PBC, BC and BS using reciprocating friction tester of pin on disk type. From this study, the result was shown that the AlBC and PBC with alloy elements were excellent to resistance wear. As the sliding speed was increased, the wear loss of PBC decreased than another Cu alloy.