• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.919-924
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• 2005

Fretting-wear caused by turbulence excitation for KSNP(Korea standard nuclear power plant) steam generator is investigated numerically. Secondary sides density and normal velocity are obtained by the thermal-hydraulic data of the steam generator. Because nonlinear finite element analysis is complex and time consuming, work rate is estimated by using linear analysis for simple straight 2-span tube. Wear volume and depth by using work rate calculation are estimated. Span length, secondary side fluid density and normal velocity are adopted to study the effects on the fretting-wear by turbulence excitation. When secondary sides density and normal velocity is increased, It turns out that secondary side density and normal gap velocity are very important paramater for fretting-wear phenomena of the steam generator.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.493-498
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• 2002

The letdown system of pressurized water reactor (PWR) nuclear fewer plants had experienced instabilities in letdown system due to unacceptable flow characteristics of control valves. The Korean Standard Nuclear Power Plants (KSNPs) have three flow paths in parallel for letdown new control. Each flow path consists of two offices and one isolation valve. This study evaluates the effect of orifice arrangement and valve stroke time of letdown isolation valve on the system transients because sudden flow changes due to valve actuation can generate high pressure peaks in letdown line. A pressure transient analysis has been preformed to evaluate the impact of dynamic transients. This analysis uses MMS which is a simulation code developed by EPRI based on the method of characteristics. The result shows that the pressure peak is reduced in the continuous arrangement but negligible. Additionally, it shows that the stroke time of linear type flog valve greater than 15 seconds can give more stable performance.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.536-547
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• 1994

The state feedback optimal control techniques are used in designing the reactor control system. The mathematical plant model with the temperature feedback effects is established from the one delayed neutron group point kinetics equation and the singly lumped thermal-hydraulic balance equations, and is expressed in terms of state variables. The LQR (Linear Quadratic Regulator) control system is designed, being followed by the LQG (Linear Quadratic Gaussian) design to determine the optimal conditions of rod movement for the desired reactor power responses. And two different servo control schemes, the ordinary feedback system and the order increased regulating system, are proposed for the purpose of input tacking. The general control characteristics such as stability margins and output responses are discussed. Comparing each other, it is found that the order increased regulating system has far better control characteristics than the ordinary feedback system.

• Geomechanics and Engineering
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• v.16 no.1
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• pp.21-34
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• 2018

Discontinuities considerably affect the mechanical and hydraulic properties of rock mass. These properties of the rock mass are influenced by the geometry of the discontinuities to a great extent. This paper aims to render an account of the geometrical parameters of several discontinuity sets related to the surrounding rock mass of Rudbar Lorestan Pumped Storage Power Plant powerhouse cavern making use of the linear and areal (circular and rectangular) sampling methods. Taking into consideration quite a large quantity of scanline and the window samplings used in this research, it was realized that the areal sampling methods are more time consuming and cost-effective than the linear methods. Having corrected the biases of the geometrical properties of the discontinuities, density (areal and volumetric) as well as the linear, areal and volumetric intensity accompanied by the other properties related to four sets of discontinuities were computed. There is an acceptable difference among the mean trace lengths measured using two linear and areal methods for the two joint sets. A 3D discrete fracture network generation code (3DFAM) has been developed to model the fracture network based on the mapped data. The code has been validated on the basis of numerous geometrical characteristics computed by use of the linear, areal sampling methods and volumetric method. Results of the linear sampling method have significant variations. So, the areal and volumetric methods are more efficient than the linear method and they are more appropriate for validation of 3D DFN (Discrete Fracture Network) codes.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.8
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• pp.408-418
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• 2017

Supercritical carbon dioxide ($sCO_2$) power system is emerging technology because of its high cycle efficiency and compactness. Meanwhile, PCHE (Printed Circuit Heat Exchanger) is gaining attention in $sCO_2$ power system technology because PCHE with high pressure-resistance and larger heat transfer surface per unit volume is fundamentally needed. Thermo-fluidic characteristics of $sCO_2$ in the micro channel of PCHE should be investigated. In this study, heat transfer characteristics of $sCO_2$ of various inlet conditions and cross-sectional shapes of single micro channel were investigated experimentally. Experiment was conducted at supercritical state of higher than critical temperature and pressure. Test sections were made of copper and hydraulic diameter was 1 mm. Convective heat transfer coefficients were measured according to each interval of the channel and pressure drop was also measured. Convective heat transfer coefficients from experimental data were compared with existing correlation. In this study, using measured data, a new empirical correlation to predict near critical region heat transfer coefficient is developed and suggested. Test results of single channel will be used for design of PCHE.

• Journal of Energy Engineering
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• v.23 no.2
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• pp.149-157
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• 2014

The steam generator tubes of nuclear power plants have various types of corrosion failures during the plant operation. The stress corrosion cracking which occurs on the outer surface of tube is called the secondary side stress corrosion cracking and mainly occurs in the expansion-transition area of tube. The causes are the concentration of impurities by the sludge pile-up related to the geometry of its region and the residual stress by tube expansion in the process of steam generator manufacturing. Especially the directionality and sizes of residual stresses are differed according to the tube expansion methods and the direction and the frequency of tube cracks depend on their characteristics. In bases on the plant experiences, it is notified that circumferential cracks of tubes expanded with explosive expansion method are dominantly occurred compared to those of tubes done with hydraulic expansion one. Therefore in this study, according to tube expansion methods frequencies and sizes of tube cracks with specific direction are compared by means of accelerated immersion test and also the crack morphology and the specific chemicals from water-chemistry environment are observed through the fracture surface examination.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.18 no.9
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• pp.937-942
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• 2008

This paper presents a case history on failures of impeller and shaft due to pressure pulsation at single stage feed water pumps in 700 MW nuclear power plant during commissioning operation. The pumps had been service and had run for approximately $40{\sim}50$ hours. For the most part, the failures of impeller occurred with the presence of a number of fatigue cracks. All cracks were associated with the deleterious surface layer of impeller by visual and metallurgical examination. On-site testing and analytical approach was performed on the systems to diagnose the problem and develop a solution to reduce the effect of exciting sources. A major concern at high-energy centrifugal pump is the pressure pulsation created from trailing edge of the Impeller blade, flow separation and recirculation at centrifugal pumps of partial load. Pressure pulsation due to the interaction generating between impeller and casing coincided with natural frequencies of the impeller and shaft system during 1ow load operation. It was identified that dynamic stress exceeding the fatigue strength of the material at the thin shroud section due to the hydraulic instability at running condition below BEP.

• Journal of the Korea Convergence Society
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• v.10 no.12
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• pp.293-298
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• 2019

The global market for temperature sensors for power plants is estimated at around 35 billion won, of which South Korea relies on imported products for more than 95 percent. This study is that a temperature measurement device for gas turbine rotators for power plants and can be applied to more than 800 of 100 MW gas turbine generators operating in Korea. This study has improved durability by changing the shape of the measuring part, structure of the connecting part, and material changes, and is a component technology applicable to other measuring devices such as humidity, gas and hydraulic pressure used in precision chemical process and plant export industry. As a result of this study, temperature sensors designed as three types of sensors for measuring the temperature of the gas turbine for power plants met Class 1 temperature accuracy in the range of 0℃ to 300℃, and improved durability significantly compared to similar products.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.315-334
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• 2014

Wolsong Low- and Intermediate-level radioactive waste (LILW) disposal center has two different types of disposal facilities and interacts with the neighboring Wolsong nuclear power plant. These situations impose a high level of complexity which requires in-depth understanding of phenomena in the safety assessment of the disposal facility. In this context, multidimensional radionuclide transport model and hydraulic performance assessment model should be developed to identify more realistic performance of the complex system and reduce unnecessary conservatism in the conventional performance assessment models developed for the $1^{st}$ stage underground disposal. In addition, the advanced performance assessment model is required to calculate many cases to treat uncertainties or study parameter importance. To fulfill the requirements, this study introduces the development of two-dimensional integrated near-field performance assessment model combining near-field hydraulic performance assessment model and radionuclide transport model for the $2^{nd}$ stage near-surface disposal. The hydraulic and radionuclide transport behaviors were evaluated by PORFLOW and GoldSim. GoldSim radionuclide transport model was verified through benchmark calculations with PORFLOW radionuclide transport model. GoldSim model was shown to be computationally efficient and provided the better understanding of the radionuclide transport behavior than conventional model.