• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.401.1-401
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.359.2-359
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.228.2-228
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• 1999

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.422-429
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• 2013

It is of great importance to localize leakages in complex pipelines for assuring their safety. A sensor array that can detect where leakages occur enables us to monitor a wide area with a relatively low cost. Beamforming is a fast and efficient algorithm to estimate where sources are, but it is generally made use of in free field condition. In practice, however, many pipelines are placed in a closed space for the purpose of safety and maintenance. This leads us to take reflected waves into account to the beamforming for interior leakage localization. Beam power distribution of reflected waves in a closed space is formulated, and spatial average is introduced to suppress the effect of reflected waves. Computer simulations and experiments ensure how the proposed method is effective to localize leakage in a closed space for structural health monitoring.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.165-172
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• 2016

An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

• Journal of the Korea institute for structural maintenance and inspection
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• v.25 no.1
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• pp.112-119
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• 2021

The failure mode of piping systems due to seismic loads is the low-cycle fatigue failure with ratcheting, and it was found that the element in which nonlinear behavior is concentrated and damage occurs is the elbow. In this study, to quantitatively express the failure criteria for a pipe elbow of SCH40 3-inch carbon steel under low-cycle fatigue, the limit state was defined as leakage, and the in-plane cyclic loading test was conducted. For the carbon steel pipe elbow, which is the vulnerable part to seismic load of piping systems, the damage index was represented using the moment-deformation angle relationship, and it was compared and analyzed with the damage index calculated using the force-displacement relationship. An attempt was made to quantitatively express the limit state of the carbon steel pipe elbow involving leakage using the damage index, which was based on the dissipated energy caused by repeated external forces.

• Journal of Energy Engineering
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• v.11 no.1
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• pp.18-25
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• 2002

One of the most serious concern in nuclear power plant piping maintenance is thickness reduction due to flow-accelerated corrosion (FAC). Since the FAC occurs under specific conditions of pH, dissolved oxygen, temperature, flow velocity, steam quality of the fluid and materials and geometry of the piping, a systematic approach is required for managing the FAC problem. In this study, construction of a secondary piping database, analyzing the FAC rate using the database and predicting the residual life was performed for a domestic CANDU nuclear power plant. Also FAC mechanism and factors affecting FAC were reviewed. By showing a case study on analysis for a pipe line between a separator and a flash tank, a procedure for managing FAC problem is suggested. The procedure proposed in this paper can be widely applied to the secondary piping of other domestic nuclear polder plants.

• Journal of the Korean Institute of Gas
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• v.10 no.3 s.32
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• pp.7-12
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• 2006

In this paper, a new evaluation scheme is suggested to estimate load-carrying capacities of wall thinned pipes. At first, computational fluid dynamics analyses employing steady-state and incompressible flow are carried out to determine pressure distributions in accordance with conveying fluid. Then, the variational pressures are applied as input condition of structural finite element analyses to calculate local stresses at the deepest point. The efficiency of proposed scheme was proven from comparison to conventional analyses results and it is recommended to consider the fluid structure interaction effect for exact integrity evaluation.