• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.732-740
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• 2022

A piping system stress analysis need to be re-performed for structural integrity assessment after reinforcement of a pipe with significant wall thinning. For efficient stress analysis, a one-dimensional beam element for the wall-thinned pipe with reinforcement needs to be developed. To develop the beam element, this work presents analytical equations for elastic stiffness of the wall-thinned pipe with reinforcement are analytically derived for axial tension, bending and torsion. Comparison with finite element (FE) analysis results using detailed three-dimensional solid models for wall-thinned pipe with reinforcement shows good agreement. Implementation of the proposed solutions into commercial FE programs is explained.

• 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 Energy Engineering
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• v.23 no.3
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• pp.116-124
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• 2014

In the plants operated under high temperature condition, piping system load analysis is often performed to prevent accidents caused by thermal deformation and also to locate inspection prioritity points of the piping system. In this study, piping system stress analysis was performed for a pipe system between the reactors in a process plant. The piping system includes typically installed hangers and expansion joints. In order to evaluate the effects of structural components such as hangers and expansion joints, the case for the expansion joint or the hanger under abnormal operation is considered. By comparison anlaysis results of piping system during normal operation and abnormal operation, the role of each pipe components are studied.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.05a
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• pp.780-785
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• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of a nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve or heather generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 7nn nuclear power plant. The exciting sources and response or the piping system are investigated by using on site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3Hz, 4.4Hz and 6.6Hz transferred from main steam header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness were installed to reduce excessive vibration.

• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.1
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• pp.21-31
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• 2013

In this study, dynamic characteristics and seismic capacity of the nuclear power plant piping system are evaluated by model test results using multi-platform shake table. The model is 21.2 m long and consists of straight pipes, elbows, and reducers. The stainless steel pipe diameters are 60.3 mm (2 in.) and 88.9 mm (3 in.) and the system was assembled in accordance with ASME code criteria. The dynamic characteristics such as natural frequency, damping and acceleration responses of the piping system were estimated using the measured acceleration, displacement and strain data. The natural frequencies of the specimen were not changed significantly before and after the testing and the failure and leakage of the piping system was not observed until the final excitation. The damping ratio was estimated in the range of 3.13 ~ 4.98 % and it is found that the allowable stress(345 MPa) according to ASME criteria is 2.5 times larger than the measured maximum stress (138 MPa) of the piping system even under the maximum excitation level of this test.

• KEPCO Journal on Electric Power and Energy
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• v.5 no.4
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• pp.323-330
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• 2019

The 500 MW Korean standard coal-fired power plant is the largest standardized power plant in Korea and has played a pivotal role in domestic power generation for over 20 years. In addition to the aging degradation due to long term operation, the probability of failure of power generation facilities is increasing due to frequent startup and stop caused by the lower utilization rate due to air pollution problem caused by coal-fired power plants. Among them, steam piping plays an important role in transferring high-temperature & pressure steam produced in a boiler to turbine for power generation. In recent years, failure of steam piping of large coal-fired power plant has frequently occurred. Therefore, in this study, failure analysis of high pressure piping weld was conducted. We identify the damage caused by high stress due to abnormal supporting structure of the piping and suggest improved supporting structure to eliminate high stress through microstructure analysis and piping stress analysis to prevent the occurrence of the similar failure of other power plant in the case of repetitive damage to the main steam piping system of the 500 MW Korean standard coal-fired power plant.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.432-437
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• 2000

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The the radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.56-63
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• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• Structural Engineering and Mechanics
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• v.72 no.1
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• pp.31-41
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• 2019

Structural integrity assessment of piping components is of paramount important for remaining life prediction, residual strength evaluation and for in-service inspection planning. For accurate prediction of these, a reliable fracture parameter is essential. One of the fracture parameters is stress intensity factor (SIF), which is generally preferred for high strength materials, can be evaluated by using linear elastic fracture mechanics principles. To employ available analytical and numerical procedures for fracture analysis of piping components, it takes considerable amount of time and effort. In view of this, an alternative approach to analytical and finite element analysis, a model based on relevance vector machine (RVM) is developed to predict SIF of part through crack of a piping component under fatigue loading. RVM is based on probabilistic approach and regression and it is established based on Bayesian formulation of a linear model with an appropriate prior that results in a sparse representation. Model for SIF prediction is developed by using MATLAB software wherein 70% of the data has been used for the development of RVM model and rest of the data is used for validation. The predicted SIF is found to be in good agreement with the corresponding analytical solution, and can be used for damage tolerant analysis of structural components.