• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.27-34
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• 2011

The objective of this study is to classify the geometry of wall-thinning defect that causes a circumferential crack in the pipe elbows subjected to internal pressure. For this objective, first of all a criterion to determine the occurrence of circumferential cracking at wall-thinned area was developed based on finite element simulation for burst tests of pipe elbow specimens that showed axial and circumferential cracking at wall-thinned area. In addition, parametric finite element analysis including various wall-thinning geometries, locations, and pipe geometries was conducted and the wall-thinning geometries that initiate circumferential crack were determined by applying the criterion to the results of parametric analysis. It showed that the circumferential crack occurs at wall-thinning defect, which has a deep, wide, and short geometry. Also, it is indicated that the pipe elbows with larger radius to thickness ratio are more susceptible to circumferential cracking at wall-thinned area.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.1-6
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.31-37
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• 2017

The heat transfer function or thermal performance is the most important function of the steam generator component in nuclear power plants. The declining of thermal performance, fouling does not affect the electric power of the nuclear power plant within a certain fouling level, but it affects the output when goes beyond the governor valve wide open of the turbine. The VWO steam pressure can be predicted through the thermal performance evaluation of steam generators in the nuclear power plant. In consideration of the fouling characteristics of the steam generator, methods of the thermal performance evaluation and fouling cases are reviewed, and also the critical VWO value is estimated through the actual thermal performance evaluation. It is necessary to apply the VWO theory based on the thermal performance of the steam generators.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.48-53
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• 2017

We diagnosed the defect using the data obtained from the nuclear power plant simulator. In this paper, we diagnosed faults in the nuclear power plant system for discovery instead of the traditional single-component or device unit. We created the six fault scenarios and used a fault simulator to obtain the fault data. It was extracted pattern from acquired failure data. Neural network model was trained and simple pattern matching algorithm was applied. We presented a simulation result and confirmed that the applied algorithm works correctly.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2631-2636
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• 2008

The most components and piping of the secondary side of domestic nuclear power plants were manufactured carbon-steel and low-alloy steel. Flow accelerated corrosion leads to wall thinning (metal loss) of carbon steel components and piping exposed to the flowing water or wet steam of high temperature, pressure, and velocity. The feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which increases as operating time progress. Several nuclear power plants in Korea have also experienced wall thinning damage in the shell wall around the impingement baffle. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the experimental results based on down-scaled experimental facility. The experiments were performed based on several types of impingement baffle plates which are installed in low pressure feedwater heater.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.1-5
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• 2023

This paper describes a newly-developed analysis methodology in comprehensive vibration assessment program (CVAP) of reactor internals to develop a valid-prototype for the design of nuclear power plants. The new analysis methodology developed in this study will be confirmed through a scale model testing (SMT). Based on the measurements obtained from dynamic pressure transducers in the SMT, a new non-dimensional equation is developed to apply the forcing functions at reactor internals for the prototype. In addition to the new non-dimensional equation, a computational fluid dynamics(CFD) is used to develop the application of the hydraulic loads at reactor internals for the prototype.

• Korean Journal of Agricultural Science
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• v.38 no.3
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• pp.493-504
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• 2011

This study was carried out to safety evaluation, the practical application and improvement of design method of the agricultural reservoir due to raising embankments. Also, it was to compare and analyze the pore water pressure (PWP), seepage (leakage) quantity and piping phenomenon according to high water table and rapid drawdown. The seepage analysis by finite element analysis was used for steady state and transient condition. The pore water pressure distribution for steady state and transient condition showed positive(+) PWP on the upstream slope, it was gradually changed negative(-) PWP on the downstream slope. The PWP in the core ranged from 100 ~ -33 kPa, the seepage line in the incline-type core suddenly decreased towards the lower levels from the higher levels. The PWP according to rapid drawdown is remained in the vicinity of the upstream slope, therefore, it is investigated to be in an unstable state by the slope stability analysis. The PWP after raising embankments showed smaller than in the before raising embankments. It was likely to be the piping phenomenon because the gradients in the before raising embankments showed largely at downstream slope, and the stability for piping in the after raising embankments increased stable state. The seepage quantity per 1 day and the leakage per 100m for the steady state and transient condition appeared to be safe against the piping. It reduced slightly regardless of the transient condition before the raising embankments and it decreased largely about 2.4 times in the early days after the raising embankments.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.28-35
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• 2021

The flow accelerated corrosion (FAC) is one of significant aging and degradation mechanism and can affect structural integrity of CANDU feeder pipes. Pipe burst can occur under normal operation pressure (min. 10 MPa) if wall-thinning of the feeder pipe due to FAC is accumulated. Previous studies considered simple shapes of feeder pipe with local wall-thinning in order to conservatively assess structural integrity of wall-thinned feeder pipe. In this paper, a new FE model is developed, having an actual shape of the feeder pipe (double bent) as well as the actual wall-thinning shape and location based on the in-service inspection result. Then, the burst pressure assessment of the wall-thinned feeder pipe is performed using lower bound limit load analysis considering elastic-perfectly plastic material. In addition, an improved formulation to predict the burst pressure of the wall-thinned feeder pipe is presented and the safety margin is compared with an existing assessment method.

• Journal of The Korean Society of Agricultural Engineers
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• v.60 no.1
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• pp.1-10
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• 2018

This study analyzed pore water pressure, seepage and leakage quantity, height of seepage and critical hydraulic gradient in order to suggest the seepage characteristics of agricultural reservoir embankment considering filter interval. The seepage characteristics of a deteriorated reservoir embankments were conducted according to the horizontal filter intervals range using three- dimensional finite element analysis. The wider the horizontal filter interval, the higher the pore water pressure increased, and the pore water pressure ratio in the center of the core has a greater effect than the base part. The seepage and leakage quantity appeared largely in the two-dimensional analysis conditions (case 1), where the filter was constructed totally in the longitudinal direction of the embankment, the wider the horizontal filter interval was gradually reduced. The reasonable filter intervals to yield efficient seepage characteristics were within 30 m for the pore water pressure of the core and the height of the seepage line. The stability of the filter installation was able to evaluate the stability of the piping by the critical hydraulic gradient method. The deteriorated reservoir with no filters or decreased functionality can significantly reduce the possibility of piping by simply installing a filter on the downstream slope. In the future, the deteriorated reservoir embankment should be checked for the reservoir remodeling because the core and filter functions have been lost or decreased significantly. In the case of a new installation, the seepage characteristic behavior due to the core and filter changes should be applied to the field after obtaining a reasonable horizontal filter interval that satisfies the safety factor by a three-dimensional analysis.