• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.14 no.1
• /
• pp.58-65
• /
• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• Transactions of the Korean Society of Automotive Engineers
• /
• v.21 no.6
• /
• pp.1-7
• /
• 2013

This paper focuses on cooling fan control by using a magnetic clutch type for the improvement of fuel economy on a heavy city bus. In general, Heavy duty vehicles use viscous clutch type cooling fan which has some disadvantages, such as slow response, wide temperature variation of engine coolant water. But a magnetic clutch type cooling fan can be controlled electronically so the engine coolant temperature can be precisely controllable and this effects could be used to reduce fuel consumption. A control system for applying the magnetic clutch type cooling fan was developed in this study and applied to the real field test and chassis dynamometer test. The result showed well controlled coolant temperature and enhancement of fuel economy.

• Journal of the Korean Society of Safety
• /
• v.33 no.3
• /
• pp.92-98
• /
• 2018

The flow of cooling water in a passive containment cooling system (PCCS), used to remove heat released in design basis accidents from a concrete containment of light water nuclear power plant, was conducted in order to investigate the thermo-fluid equilibrium among many parallel tubes of PCCS. Numerical simulations of the subcooled boiling flow within a coolant loop of a PCCS, which will be installed in innovative pressurized-water reactor (PWR), were conducted using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. As the first step, the simplified geometry of PCCS with 36 tubes was modeled in order to reduce computational resource. Even and uneven thermal loading conditions were applied at the outer walls of parallel tubes for the simulation of the coolant flow in the PCCS at the initial phase of accident. It was observed that the natural circulation maintained in single-phase for all even and uneven thermal loading cases. For uneven thermal loading cases, coolant velocity in each tube were increased according to the applied heat flux. However, the flows were mixed well in the header and natural circulation of the whole cooling loop was not affected by uneven thermal loading significantly.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1689-1696
• /
• 2020

In this study, we developed a fiber-optic humidity sensor (FOHS) system for the monitoring and detection of coolant leakage in nuclear power plants. The FOHS system includes an FOHS, a spectrometer, a halogen white-light source, and a Y-coupler. The FOHS is composed of a humidity-sensing material, a metal tube, a multi-mode plastic optical fiber, and a subminiature version A (SMA) fiber-optic connector. The humidity-sensing material is synthesized from a mixture of polyvinylidene fluoride (PVDF) in dimethyl sulfoxide (DMSO) and hydroxyethyl cellulose (HEC) in distilled water. We measured the optical intensity of the light signals reflected from the FOHS placed inside the humidity chamber with relative humidity (RH) variation from 40 to 95%. We found that the optical intensity of the sensing probe increased linearly with the RH. The reversibility and reproducibility of the FOHS were also evaluated.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.358-363
• /
• 1995

A study on passive cooling systems for concrete containment of advanced pressurized water reactors has been performed. The proposed passive containment cooling system (PCCS) consist of (1) condenser units located inside containment, (2) a steam condensing pool outside containment at higher elevation, and (3) downcommer/riser piping systems which provide coolant flow paths. During an accident causing high containment pressure and temperature, the steam/air mixture in containment is condensed on the outer surface of condenser tubes transferring the heat to coolant flowing inside tubes. The coolant transfers the heat to the steam condensing pool via natural circulation due to density difference. This PCCS has the following characteristic: (1) applicable to concrete containment system, (2) no limitation in plant capacity expansion, (3) efficient steam condensing mechanism (dropwise or film condensation at the surface of condenser tube), and (4) utilization of a fully passive mechanism. A preliminary conceptual design work has been done based on steady-state assumptions to determine important design parameter including the elevation of components and required heat transfer area of the condenser tube. Assuming a decay power level of 2%, the required heat transfer area for 1,000MWe plant is assessed to be about 2,000 ㎡ (equivalent to 1,600 of 10 m-long, 4-cm-OD tubes) with the relative elevation difference of 38 m between the condenser and steam condensing pool and the riser diameter of 0.62 m.

• Journal of the Mineralogical Society of Korea
• /
• v.18 no.2
• /
• pp.117-125
• /
• 2005

It is common to find lots of organic and inorganic precipitates inside of industrial cooling system. Analytical instruments including ICP-AES, IC, XRD, and SEM were used to investigate the effects of application of Mg on reaction among coolant, precipitates and suspended matters within chiller system. Magnesium (Mg) has a decreasing effect on total dissolved material in the coolant suggesting a significant improvement of quality of coolant. Disappearance of most organic materials in the cooling devices also suggests an inhibitive effect on the growth of organic matters.

• Transactions of the Korean Society of Automotive Engineers
• /
• v.19 no.6
• /
• pp.53-60
• /
• 2011

The effect of exhaust heat recovery system can be evaluated by two well known method. First method is to measure the time duration from engine start under cold coolant temperature till coolant get warmed. By this methodology coolant warming duration can be index of warm-up effect. Second method is to analyze heat balance of the engine during warm-up phase under steady engine operation so that wasted energy by losses such as cooling and exhaust can be index of warm-up effect. This study focused on evaluation of warming-up effect by both methodology above mentioned using 2L SI engine under from idle to 2000rpm steady condition. Results, idle operation showed low heat recovery efficiency but under higher engine speed condition, remarkable heat recovery efficiency improvement was observed. In 2000rpm steady condition, warm-up duration of engine is decreased by exhaust heat recovery system.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.16 no.3
• /
• pp.16-23
• /
• 2017

For most machine tools, it is necessary to remove chips and coolant oil because it they will continue to be created during the manufacture of workpieces. Existing products that are in use are installed and used as they reflect depending on the characteristics of each device separately. This study proposes a method to remove the security chip as well as developing an integrated system capable of reducing coolant damage. The Leverage AutoCAD and CATIA program was used for 2D and 3D design, shapes were identified by utilizing the KeyShot program, and the load and displacement analysis of the development apparatus was performed utilizing the ANSYS program. After the prototype underwent sufficient design review, the mixed oil separation device had a complete sensor control program using the LabVIEW program. The chip design process for transferring experiments and experiments on the mixed oil cooling device were developed for performance tests of the product. The final product resulted in an increase in space utilization during commercialization, reduced installation costs, and caused social effects such as pulmonary flow reduction, which, through the economic costs, reduces pollution, resulting in various benefits to the industry, such as deceased errors in the workplace decreases.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1540-1553
• /
• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• The KSFM Journal of Fluid Machinery
• /
• v.3 no.2 s.7
• /
• pp.24-30
• /
• 2000

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seen that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stress predictions.