• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.492-500
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• 2004

The purpose of this study is localization of safety relief valves fur Nuclear Service through technical development with overall design, fabrication, inspection, capacity certification test and functional qualification test of safety relief valves in accordance with KEPIC MN Code(or ASME Sec.III ). The safely relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. But we're depending on technology of the other country up to the present time. Because we don't have our own technologies, we have been spent the great time and money on installing and repairing safety relief valve at nuclear power plant. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Journal of Mechanical Science and Technology
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• v.19 no.5
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• The KSFM Journal of Fluid Machinery
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• v.8 no.3 s.30
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• pp.16-25
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• 2005

The purpose of this study is localization of safety relief valves for Nuclear Service. The safety relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. We developed design technology used FEM ' CFM about safety relief valve for Nuclear Service according to ASME (or KEPIC) Code and KHNP's Technical Specification. To prove validity of a design technology, actually, we manufactured and inspected and tested the sample products designed according to a developed technology. The capacity qualification test was achieved according to requirement of ASME(or KEPIC) Code by NBBI and the functional qualification test was achieved according to ASME QME-1 for operating condition in technical specification of KHNP by NLI. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.835-838
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• 2002

An experiment has been performed using a facility, which simulates the safety depressurization system (SDS) and in-containment refueling water storage tank (IRWST) of APR1400, an advanced PWR being developed in Korea, to investigate the dynamic load resulting from the blowdown of steam from a steam generator through a sparser. The influence of the key parameters, such as air mass, steam pressure, submergence, valve opening time, and pool temperature, on frequency and peak toads was investigated. The blowdown phenomenon was analyzed to find out the real cause of the initiation of bubble oscillation and discrepancy in frequencies between the experiment and calculation by conventional equation for bubble oscillation. The cause of significant damping was discussed and is presumed to be the highly tortuous flow path around bubble. The Rayleigh-Plesset equation, which is modified by introducing method of image, reasonably reproduces the bubble oscillation in a confined tank. Right after the completion of air discharge the steam discharge immediately follows and it condenses abruptly to provide low-pressure pocket. It may contribute to the negative maximum being greater than positive maximum. The subsequently discharging steam does not play as at the driving force anymore.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.7-12
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• 2000

The automatic turbine startup system provides turbine control based on thermal stress. During the startup, control system monitors and evaluates main components of turbine using damage mechanism and life assessment. In case of valve chest, the temperature of inner/outer wall is measured by thermo-couples and the safety of these values are evaluated by using allowable ${\Delta}T$ limit curve during the startup. Because allowable ${\Delta}T$ limit curve includes life assessment, it is possible to apply this curve to turbine control system. In this paper, low cycle fatigue damage and combined rupture and low cycle fatigue damage criterion proposed for yielding the allowable ${\Delta}T$ limit curve of CV(control valve) chest. To calculate low cycle fatigue damage, the stress analysis of valve chest has peformed using FEM. Automatic turbine startup to assure service life of CV was achieved using allowable ${\Delta}T$ limit curve.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.1-6
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• 2002

The automatic turbine startup system provides turbine control based on thermal stress. During the startup, control system monitors and evaluates main components of turbine using damage mechanism and life assessment. In case of valve chest, the temperature of inner/outer wall is measured by thermo-couples and the safety of these values are evaluated by using allowable △T limit currie during the startup. Because allowable ΔT limit curve includes life assessment, it is possible to apply this curve to turbine control system. In this paper, low cycle fatigue damage, combined rupture and low cycle fatigue damage criterion were proposed for yielding the allowable ΔTf limit curve of CV(control valve) chest. To calculate low cycle fatigue damage, the stress analysis of valve chest has been performed using FEM. Automatic turbine startup to assure service life of CV was achieved using allowable ΔT limit curve.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2173-2182
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• 2020

Control valves are widely used to regulate fluid flux in nuclear power plants, and there are more than 1500 control valves in the primary circuit of one nuclear power plant. With their help, the flux can be regulated to a specific level of water or steam to guarantee the energy efficiency and safety of the nuclear power plant. The flux characteristics of the control valve mainly depend on the valve core shape. In order to analyze the effects of valve core shapes on flux characteristics of control valves, this paper focuses on the valve core shapes. To begin with, numerical models of different valve core shapes are established, and results are compared with the ideal flux characteristics curve for the purpose of validation. Meanwhile, the flow fields corresponding to different valve core shapes are investigated. Moreover, relationships between the valve core opening and the outlet flux under different valve core shapes are carried out. The flux characteristics curve and equation are proposed to predict the outlet flux under different valve core openings. This work can benefit the further research of the flux control and the optimization of the valve core for control valves in nuclear power plants.