• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Journal of the Korean Society of Safety
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• v.33 no.5
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• pp.164-169
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• 2018

A detailed fatigue evaluation procedure was developed to mitigate the excessive conservativeness of the conventional environmental fatigue evaluation method for the pressurizer spray line elbow of domestic new nuclear power plants. The pressurizer spray line is made of austenitic stainless steel, which is relatively sensitive to the environmentally assisted fatigue, and has a low degree of design margin in terms of environmentally assisted fatigue due to the thermal stratification phenomenon on the pipe cross section as a whole or locally. In this study, to meet the environmental fatigue design requirements of the pressurizer spray line elbow, the new environmental fatigue evaluation has been performed, which used the ASME Code NB-3200-based detailed fatigue analysis and the environmental fatigue correction factor instead of the existing NB-3600 evaluation method. As a result, the design requirements for environmentally assisted fatigue were met in all parts of the pressurizer spray line elbow including the fatigue weakened zones by thermal stratification.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.58-65
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• 2016

Since, in case of high energy piping, steam jets ejected from the rupture zone may cause damage to nearby structure, it is necessary to design it into consideration of nuclear power plant design. For the existing nuclear power plants, the ANSI / ANS 58.2 technical standard for high-energy pipe rupture was used. However, the US Nuclear Regulatory Commission (USNRC) and academia recently have pointed out the non-conservativeness of existing high energy pipe fracture evaluation methods. Therefore, it is necessary to develop a highly reliable evaluation methodology to evaluate the behavior of steam jet ejected during high energy pipe rupture and the effect of steam jet on peripheral devices and structures. In this study, we develop a method for analyzing the impact load of a jet by high energy pipe rupture, and plan to carry out an experiment to verify the evaluation methodology. In this paper, the basic data required for the design of the jet impact load experiment equipment under construction, 1) the load change according to the jet distance, 2) the load change according to the jet collision angle, 3) the load variation according to structure diameter, and 4) the load variation depending on the jet impact position, are numerically obtained using the developed steam jet analysis technique.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.5
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• pp.466-472
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• 2009

An acoustic leak monitoring system(ALMS) using acoustic emission(AE) technique was applied for leakage detection of nuclear power plant's pipeline which is operated in high temperature and pressure condition. Since this system only monitors the existence of leak using the root mean square(RMS) value of raw signal from AE sensor, the difficulty occurs when the characteristics of leak size and shape need to be evaluated. In this study, dual monitoring system using AE sensor and accelerometer was introduced in order to solve this problem. In addition, artificial neural network(ANN) with Levenberg.Marquardt(LM) training algorithm was also applied due to rapid training rate and gave the reliable classification performance. The input parameters of this ANN were extracted from varying signal received from experimental conditions such as the fluid pressure inside pipe, the shape and size of the leak area. Additional experiments were also carried out and with different objective which is to study the generation and characteristic of lamb and surface wave according to the pipe thickness.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.77-87
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• 2018

In the case of high-energy line breaks in nuclear power plants, supersonic steam jet is formed due to the rapid depressurization. The steam jet can cause impingement load on the adjacent structures, piping systems and components. In order to secure the design integrity of the nuclear power plant, it is necessary to evaluate the load characteristics of the steam jet generated by high-energy pipe rupture. In the design process of nuclear power plant, jet impingement load evaluation was usually performed based on ANSI/ANS 58.2. However, U.S. NRC recently pointed out that ANSI/ANS 58.2 oversimplifies the jet behavior and that some assumptions are non-conservative. In addition, it is recommended that dynamic analysis techniques should be applied to consider transient load characteristics. Therefore, it is necessary to establish an evaluation methodology that can analyze the dynamic load characteristics of steam jet ejected when high energy pipe breaks. This research group has developed and validated the CFD analysis methodology to evaluate the transient behavior of supersonic impinging jet in the previous study. In this study, numerical study on the transient load characteristics of supersonic steam jet impingement was carried out and amplitude and frequency analysis of transient jet load was performed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.8
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• pp.668-677
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• 2008

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which consideded safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

• Journal of the Korean Society for Nondestructive Testing
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• v.32 no.2
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• pp.170-176
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• 2012

In the nuclear industry, wall thinning defect of straight pipe occur the enormous loss in life evaluation and safety evaluation. To use non-destructive technique, we measure deformation, vibration, defect evaluation. But, this techniques are a weak that is the measurement of the wide area is difficult and the time is caught long. In the secondary side of nuclear power plants mostly used steel pipe, artificiality wall thinning defect make in the side and different thickness make to the each other, wall thinning defect part of deformation measure by using shearography. In addition, optical measurement through deformation, vibration, defect evaluation evaluate pipe and thickness defects of pressure vessel is to evaluate quantitatively. By shearography interferometry to measure the pipe's internal wall thinning defect and the variation of pressure use the proposed technique, the quantitative defect is to evaluate the thickness of the surplus. The amount of deformation use thickness of surplus prediction of the actual thickness defect and approximately 7 percent error by ensure reliability. According to pressure the amount of deformation and the thickness of the surplus through DB construction, nuclear power plant pipe use wall thinning part soundness evaluation. In this study, pressure vessel of thickness defect measure proposed nuclear pipe of wall thinning defect prediction and integrity assessment technology development. As a basic research defected theory and experiment, pressure vessel of advanced stability and soundness and maintainability is expected to contribute foundation establishment.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Journal of the Korean Society for Advanced Composite Structures
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• v.3 no.3
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• pp.1-8
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• 2012

Recently, underground pipes are utilized in various fields of applications such as sewer lines, drain lines, water mains, gas lines, telephone and electrical conduits, culverts, oil lines, etc. Most of pipes are installed for long-term purposes and they should be safely installed in consideration of installation conditions because there are unexpected various terrestrial loading conditions. In this paper, we present the result of investigation pertaining to the structural behavior of glass fiber reinforced thermosetting polymer plastic (GFRP) flexible pipes buried underground. The mechanical properties of the GFRP flexible pipes produced in the domestic manufacturer are determined and the results are reported in this paper. In addition, ring deflection is measured by the field tests and the finite element analysis (FEA) is also conducted to simulate the structural behavior of GFRP pipes buried underground. From the field test results, we predicted long-term, up to 50 years, ring deflection of GFRP pipes buried underground based on the method suggested by the existing literature. It was found that the GFRP flexible pipe to be used for cooling water intake system in the nuclear power plant is appropriate because 5% ring deflection limitation for 50 years could be satisfied.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1164-1171
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• 2020

T-junction structures play an important role in nuclear power plant systems. Research on liquid entrainment is mostly based on small-scale branch pipes (d/D ≤ 0.2) and attention paid to large-scale branch pipes (0.33 < d/D < 1) is insufficient. Accordingly, this study implements a series of experiments on the liquid entrainment of T-junction with different angles (32.2°,47.9°,62.3°,90°) through a large-scale branch (d/D = 0.675). The onset liquid entrainment is related to the gas phase Froude number Fr_g, the dimensionless gas chamber height h_b/d and the branch pipe angle 𝜃. As Fr_g increases, h_b/d also rises. With a constant h_b/d, the onset liquid entrainment changes from droplets entrainment by the gas phase to that by the rising liquid film. The steady-state liquid entrainment is related to w_3g, h/d and 𝜃. With constant w_3g and h/d, the branch quality grows as the branch angle increases. With a certain h/d, the branch quality increases, as the w_3g number increases.