• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.10
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• pp.1612-1620
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• 2004

This study presents methods to estimate elastic-plastic crack opening displacement (COD) fur circumferential through-wall cracked pipes for the Leak-Before-Break (LBB) analysis of pressurized piping. Proposed methods are based not only on the GE/EPRI approach but also on the reference stress approach. For each approach, two different estimation schemes are given, one for the case when full stress-strain data are available and the other fur the case when only yield and ultimate tensile strengths are available. For the GE/EPRI approach a robust way of determining the Ramberg-Osgood (R-O) parameters is proposed, not only fur the case when detailed information on full stress-strain data is available but also for the case when only yield and ultimate tensile strengths are available. The COD estimates according to the GE/EPRI approach, using the R-O parameters determined from the proposed R-O fitting procedures, generally compare well with the published pipe test data. For the reference stress approach, the COD estimates according to the method based on both full stress-strain data and limited tensile properties are in good agreement with pipe test data. In conclusion, experimental validation given in the present study provides sufficient confidence in the use of the proposed method to practical LBB analyses even though when information on material's tensile properties is limited.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.307-314
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• 2004

A circumferential guided wave method was developed to detect the axial crack on the bent feeder pipe. Dispersion curves of circumferential guided waves were calculated as a function of curvature of the pipe. In the case of thin plate, i.e. infinite curvature, as the frequency increases, the $S_0$ and $A_0$ mode coincide and eventually become Rayleigh wave mode. In the case of pipe, however, as the curvature increases, the lowest modes do not coincide even in the high frequencies. Based on the analysis, a rocking technique using angle beam transducer was applied to detect an axial defect in the bent region of PHWR feeder pipe. Based on the analysis of experimenal data for artificial notches, the vibration modes of each signal were identified. It was found that the notches with the depth of )0% of wall thickness can be detected with the method.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.2
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• pp.136-141
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• 2004

A torsional guided wave mode was applied to detect a crack in a pipe. An array of electromagnetic acoustic transduce. (EMAT that can generate and receive torsional guided ultrasound with the frequency of 200kHz was designed and fabricated for testing a pipe of 2.5 inch diameter Artificial notches with various depths were fabricated in a bent feeder pipe mock-up and the detectability was examined from the distance of 2m of the specimen. The axial notches with the depth of 5% of wall thickness were successfully detected by a torsional mode (T(0,1)) generated by the EMAT However, it was found that the depth of defects was not related to the signal amplitude.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• Journal of Korea Water Resources Association
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• v.53 no.4
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• pp.303-312
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• 2020

An underground deep tunnel system is a facility in form of a reverse siphon for an under flood defense structure. In this study, the 'Shinwol rainwater storage and drainage system', which is under construction for the first time in South Korea, in order to confirm the effects of undular bore and pressurized air on the hydraulic stability of the facility in various flood scenarios a hydraulic model experiment was performed. As a result of this study, it was analyzed that the undular bore generated downstream pushed the pressurized air collected in the facility while moving upstream, and the pressure inside the pipe increased at this time. It was analyzed that the pressure during the passage of the undular bore was greater than the sum of the static pressure and dynamic pressure at the time and overflow occurred when the cross-sectional size of the pressurized air was more than 40% of the cross sectional area of the tunnel. It is determined that this is correlated with the volume of pressurized air collected in the facility, and it is determined that it is necessary to study the relationship between velocity of undular bore and the volume of pressurized air in the future.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Journal of Korean Society of Environmental Engineers
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• v.27 no.10
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• pp.1065-1071
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• 2005

Many techniques of leak detection in pipeline systems have developed based on the propagation wave speeds and wave attenuation. In this paper, the transient analysis methodology is used for calculating the wave speed in the plastic pipe and a frequency analysis methodology is developed for leakage detection in water pipe networks. Data acquisition system for pressurized pipeline system were designed md fabricated to obtain high frequency pressure data. The methodology properly handles the unavoidable uncertainties in measurement and modeling error. Based on information from head pressure test data, it provides leak prediction capability from the transient events with leakage.

• Journal of ILASS-Korea
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• v.18 no.1
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• pp.9-15
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• 2013

Wall thinning of pipeline in power plants occurs mainly by flow acceleration corrosion (FAC), cavitation erosion (C/E), liquid droplet impingement erosion (LDIE). Wall thinning by FAC and C/E has been well investigated; however, LDIE in plant industries has rarely been studied due to the experimental difficulty of setting up a long injection of highly-pressurized air. In this study, we designed a long-term experimental system for LDIE and investigate the behavior of LDIE for three kinds of materials (A106B, SS400, A6061). The main control parameter was the air-water ratio (${\alpha}$), which was defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). In order to clearly understand LDIE, the spraying velocity (${\nu}$) of liquid droplets was controled larger then 160 m/s and the experiments were performed for 15 days. Therefore, this research focuses relation between erosion rate and air-water ratio on the various pipe-flow materials. NPP(nuclear power plant)'s LDIE prediction theory and management technique were drawn from the obtained data.