• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• 한국안전학회지
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• 제21권1호
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• pp.28-34
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• 2006

A procedure of estimating failure probability is demonstrated for a pressurized pipe of CrMo steel used at $538^{\circ}C$. Probabilistic fracture mechanics were employed considering variations of pressure loading, material properties and geometry. Probability density functions of major material variables were determined by statistical analyses of implemented data obtained by previous experiments. Distributions of the major variables were reflected in Monte Carlo simulation and failure probability as a function of operating time was determined. The creep crack growth life assessed by conventional deterministic approach was shown to be conservative compared with those obtained by probabilistic one. Sensitivity analysis for each input variable was also conducted to understand the most influencing variables to the residual life analysis. Internal pressure, creep crack growth coefficient and creep coefficient were more sensitive to failure probability than other variables.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.234-239
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• 2004

A study on the probabilistic methodology for the estimation of the remaining life of Pressurized pipelines containing active corrosion defects is presented. This reliability assessment is earned out using extream value distribution of the corroded defects instead of already published failure perssure moded like NG18 or ASME B31G. The failure probability of pipelines depends on the number of corroded defects. and it could be calculated directly as the area exceeded a defined L V(Limited Value of corrosion depth). The remaining life of pressurized pipelines can also be estimated by the PDF of extream value distribution as calculating the exceeded area with a defined failure probability.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2006년도 하계학술발표대회 논문집
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• pp.132-137
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• 2006

Supercooled type ice slurry system is hard to keep a proper supercooling degree when solution becomes supercooling state. This is the reason of the ice blockage in pipe or cooling part due to an unstable cooling state. In this study, a cooling experiment was performed to pressurized solution in a stationary state. The behaviors during the supercooled aqueous solution were investigated at fixed flow rate of brine and aqueous solution of ethylene glycol 7 mass%. Also the effect to the freezing point of supercooled aqueous solution was investigated to the different pressure 101, 202, 303, and 404 kPa. At results, the pressure of the aqueous solution in the cylinder increased the supercooling degree increased and dissolution of supercooled point decreased.

• International Journal of Safety
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• 제3권1호
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• Smart Structures and Systems
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• 제19권5호
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• pp.473-485
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• 2017

Ratcheting behavior of $90^{\circ}$ elbow piping subject to internal pressure 20 MPa and reversed bending 20 kN was investigated using experimental method. The maximum ratcheting strain was found in the circumferential direction of intrados. Ratcheting strain at flanks was also very large. Moreover, the effect of temperature on ratcheting strain of $90^{\circ}$ elbow piping was studied through finite element analysis, and the results were compared with room condition ($25^{\circ}$). The results revealed that ratcheting strain of $90^{\circ}$ elbow piping increased with increasing temperature. Ratcheting boundary of $90^{\circ}$ elbow piping was determined by Chaboche model combined with C-TDF method. The results revealed that there was no relationship between the dimensionless form of ratcheting boundary and temperature.

• Earthquakes and Structures
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• 제17권5호
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• pp.489-499
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• 2019

Elastic-plastic behavior of nuclear power plant elbow piping under seismic loads has been conducted in this study. Finite element analyses are performed using classical Bilinear kinematic hardening model (BKIN) and Multilinear kinematic hardening model (MKIN) as well as a nonlinear kinematic hardening model (Chaboche model). The influence of internal pressure and seismic loading on ratcheting strain of elbow pipe is studied by means of the three models. The results found that the predicted results of Chaboche model is maximum, closely followed by the predicted results of MKIN model, and the minimum is the predicted results of BKIN model. Moreover, comparisons of analysis results for each plasticity model against predicted results for a equivalent cyclic loading elbow component and for a simplified piping system seismic test are presented in the paper.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• 대한기계학회논문집B
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• 제23권11호
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• pp.1390-1398
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• 1999

A numerical analysis for ROT sparger of PWR(Pressurized Water Reactor) is carried out. Computation is performed to investigate the flow characteristics as the change of design factor. As the result of this study, RDT sparger's flow resistance coefficient is K=3.53 at the present design condition if engineering mar&in is considered with 20%, and flow ratio into branch pipe is $Q_s/Q_i=0.41$. Velocity distribution at exit is not uniform because of separation in branch pipe. In the change of inlet flow rate and section area ratio of branch pipe for main pipe, flow resistance coefficient is increased as $Q_s/Q_i$ decreasing, but in the change of branch angle and outlet nozzle diameter of main pipe, flow resistance coefficient is decreased as $Q_s/Q_i$ decreasing. As the change rate of $Q_s/Q_i$ is the larger, the change rate of flow resistance coefficient is the larger. The change rate of pressure loss is the largest change as section area ratio changing. The optimal design condition of sparger is estimated as the outlet nozzle diameter ratio of main pipe is $D_s/D_i=0.333$, the section area ratio is $A_s/A_i=0.2$ and the branch angle is ${\alpha}=55^{\circ}$.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.1053-1060
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• 2006

This paper consider an initially deformed state caused by the pressurized fluid flowing through the pipe at a constant velocity. When the initial forte is neglected in curved pipes, the natural frequencies are reduced as flow velocity increases. However, when the initial tension took into account, the natural frequencies are not changed with the change of the flow velocity. As the internal pipe pressure is increased the natural frequencies are also slightly increased. In free vibrational simulation of piping systems in petrochemical plants, it is necessary to calculate the initial state force due to the velocity and the pressure of the fluid flow from the equilibrium first, then the force should be included in the equation of motion of the systems to get more accurate natural frequencies. In this study, calculate the mass matrix and stiffness matrix of piping system by MATLAB