• Proceedings of the Computational Structural Engineering Institute Conference
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• 1995.10a
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• pp.61-67
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• 1995

The steam generator, one of the major components in the reactor coolant system, plays an important role in transferring the thermal energy made in the reactor during normal operation to the secondary side and producing steam to drive turbine. A hydraulic snubber system is used in order to protect the steam generator under the dynamic loading condition and to absorb the thermal expansion transmitted by the reactor coolant piping due to high temperature and pressure during normal operation. In this study, the model for a geometrical linkage system is presented to analyze the snubber stroke of the steam generator and the parameters in the snubber stroke analysis are investigated. A method to analyze lever ratio of the linkage system which is required in the process of determining the snubber stiffness value is also presented. To discuss the validation of the suggested analysis, the analysis results are compared with the measured data during the hot functional test for the standardized 1000 Mwe pressurized water reactor plant under the construction.

• The KSFM Journal of Fluid Machinery
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• v.15 no.3
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• pp.57-63
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• 2012

Effects of coolant pre-cooling and fuel pre-heating on the performance of a combined cycle using a F-class gas turbine were investigated. Coolant pre-cooling results in an increase of power output but a decrease in efficiency. Performance variation due to the fuel pre-heating depends on the location of the heat source for the pre-heating in the bottoming cycle (heat recovery steam generator). It was demonstrated that a careful selection of the heat source location would enhance efficiency with a minimal power penalty. The effect of combining the coolant pre-cooling and fuel pre-heating was also investigated. It was found that a favorable combination would yield power augmentation, while efficiency remains close to the reference value.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.10
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• pp.1274-1284
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• 1999

The Regulatory Guide 1.82 recommends an analysis of hydraulic performance for sump of ECCS (Emergency Core Cooing System) when LOCA(Loss of Coolant Accident) occurs in a nuclear power plant. The present study deals with 3-dimensional, unsteady, turbulent and two-phase flow simulation to examine the behavior of mixture of reactor coolant and debris near the floor of containment building in conjunction with appropriate assumptions. The dispersed solid model has been adjusted to the interfacial momentum transfer between reactor coolant and debris. According to the results, the counterclockwiserecirculation zone had been formed in the region between sump and connection aisle about 376 second after LOCA occurs. The debris thickness accumulated on a sump screen periodically increases or decreases up to 2000 second, afterwards its peak decreases.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.538-543
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• 2020

Boric acid is used as a coolant for pressurized-water reactors, and the degree of burnup is controlled by the concentration of boric acid. Therefore, accurate measurement of the concentration of boric acid is an important factor in reactor safety. An improved system was proposed for the accurate determination of boron concentration. A new boron-concentration measurement technique, called multisensitive detection, was developed to improve the measurement accuracy of boron meters. In previous studies, laboratory-scale experiments were performed based on different sensitivity detectors, confirming a 65% better accuracy than conventional single-detector boron meters. Based on these experimental results, an experimental system simulating the coolant-circulation environment in the reactor was constructed; accuracy analysis of the boron meter with a multisensitivity detector was performed at the actual coolant pressure and temperature. In this study, the boron concentration conversion equation was derived from the calibration test, and the accuracy of the boron concentration conversion equation was examined through a repeatability test. Through the experiment, it was confirmed that the accuracy was up to 87.5% higher than the conventional single-detector boron meter.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.361-364
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• 2003

The safe operation and reliable maintenance of nuclear power plants is one of the most fundamental and important tasks. It is known that a loose part such as a disengaged and drifting metal inside of reactor coolant systems might lead to a serious damage because of their impact on the components of the coolant system. In order to estimate the impact position of a loose par, three accelerometers attached to the wall of the coolant system have been used. These accelerometers measure the vibration of the coolant system induced by loose part impact. In the conventional analysis system, the low pass filtered version of the vibration data was used for the estimation of the position of a loose part. It is often difficult to identify the initial point of the impact signal by using just a low passed time signal because the impact wave is dispersed during propagation into the sensor. In this paper, the impact signal is analysed by use of various time frequency methods including the short time Fourier transform(STFT), the wavelet transform, and the Wigner-Vill distribution for finding a convenient way to identify the starting point of a impact signal and their advantages and limits are discussed.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.161-167
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• 1994

This paper presents a concept of a pressure tube-type water-cooled reactor without the emergency core cooling system. It adopts an innovative fuel channel design using metallic fuel matrix to improve heat transfer from fuel to moderator at loss of coolant cooling. The heat produced in the fuel is cooled by the coolant system during normal operation, but by the passive moderator system at loss of coolant cooling including the loss-of-coolant accident(LOCA). Simple analysis shows that the fuel channel temperature can be maintained within the permissible range for both normal operation and a complete LOCA.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.117-123
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• 2015

To remove heat generated during a burn-up test of nuclear fuels, the heat generation rate of nuclear fuels should be calculated accurately, and a coolant should be circulated in the test loop at an adequate flow rate. HANARO is an open pool-type reactor with an independent test loop for the burn-up test of nuclear fuels. A test rig is installed in the test loop, and a coolant is circulated through the test loop to maintain the temperature of the nuclear fuel rods within a desired temperature during an irradiation test. The components and sensors in the test rig can be broken or malfunction owing to the flow-induced vibration. In this study, a coolant flow simulation system was developed to verify and confirm the soundness of components and sensors assembled in the test rig with a high flow rate of the coolant.

• Journal of ILASS-Korea
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• v.20 no.2
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• pp.65-69
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• 2015

In this study, the thermal performance of vehicle's cooling system is experimentally investigated using the water/coolant-based $Al_2O_3$ nanofluids as working fluids. For the purpose, the water/coolant-based $Al_2O_3$ nanofluids are prepared by twostep method with gum arabic. In order to obtain the well-suspended nanofluids, the agglomerated $Al_2O_3$ nanoparticles are precipitated using centrifugal force and the experiments are performed with supernatant of them. The thermal conductivity is measured by transient hot wire method and the thermal conductivity of nanofluids is enhanced up to 4.8% as compared to that of base fluids. Moreover, the cooling performance of water/coolant-based $Al_2O_3$ nanofluids is evaluated using vehicle's engine simulator under the constant RPM condition. The results show that the cooling performance of automobile engine increases up to 5.9% using prepared nanofluids. To investigate the effect of nanofluids on exhaust gas, the $NO_x$ emission is measured during the operation with respect to time and 10.3% of $NO_x$ emission is decreased. The experimental results imply that the water/coolant-based $Al_2O_3$ nanofluids might be used as a next-generation vehicles' coolant