• The KSFM Journal of Fluid Machinery
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• v.17 no.1
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• pp.5-11
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• 2014

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.310-315
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• 1997

System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.407-410
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• 2002

A series of steam explosion experiments using real core materials of $ZrO_2$ and corium(a mixture of $ZrO_{2}\;and\;UO_{2}$) has been performed to evaluate the risk of steam explosion load in nuclear power plants. Surprisingly, spontaneous steam explosions are observed far both materials, which have been thought to be inexplosive so far. The dynamic pressure and morphology of the debris clearly indicate the evidence of an explosion. The experimental results also indicate that $ZrO_2$ is more explosive than corium.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.232-235
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• 2008

The vibration of steam turbine is caused by Mass Unbalance, Shaft Misalignment, Oil Whip and Rubbing etc. but in turbine which is normally operated and maintained, the Mass Unbalance component possesses the greatest portion. Our power plant has two steam turbines in capacity of 200MW and 135MW respectively and each turbine is supported by 6 journal bearings. However, we had many difficulties because the vibration amplitude of No 3 and 4 Bearings was high during the start-up and operation mode change of steam turbine. But, with this study, we completely solved the vibration problem caused by the mass unbalance of No 1 steam turbine. Until a recent date, No 3 and 4 bearings which support high pressure turbine for No 1 steam turbine had shown about 135${\mu}$m in vibration amplitude (sometimes it increased to 221${\mu}$m maximum. alarm: 6mils, trip: 9mils) at base load. After applying the study, they decreased to about 40${\mu}$m maximum. It is a result from that we did not change the setting value of Bearing Alignment and only changed the assembly position of internal parts in Synchro Clutch Coupling Rachet Wheel which links between high pressure turbine and low pressure turbine, and increased the internal gap and machining of the Pawl stopper surface. In the operation of steam turbine, if the vibration value increases by 1X, we should reduce the vibration of bearing by weight balancing. However, unless the vibration of bearing is declined by the balancing, we will have to disassemble and check the component and find the cause. In this study, We researched the way to lower mass unbalance that is 1X vibration component which has the greatest portion of vibration generated by steam turbine and We got good result by applying the findings of this study.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.425-429
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• 2004

In real design of the high & interim pressure turbine casing, it is one of the important things to figure out its thermal strain exactly. In this paper, with the establishment of the new concept for the heat transfer coefficient of steam that is one of the factors in analysis of the thermal stress for turbine casing, an analysis was done for one of the high & interim pressure turbine casings in operating domestically. The sensitivity analysis of the heat transfer coefficient of steam to the thermal strain of the turbine casing was done with a 2-D simple model. The analysis was also done with switching of the material properties of the turbine casing and resulted in that the thermal strain of the turbine casing was not so sensitive to the heat transfer coefficient of steam. On the basis of this, 3-D analysis of the thermal strain for the high and interim pressure turbine casing was done.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.57-62
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• 2001

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named Test for Real cOrium Interaction with water (TROI) using reactor material to investigate whether the corium would lead to energetic steam explosion when interacted with cold water at low pressure. The melt-water interaction is confined in a pressure vessel with the multi-dimensional fuel and water pool geometry. The cold crucible technology, where the mixture of powder in a water-cooled cage is heated by high frequency induction, is employed. In this paper, results of the first series of tests ($TROI-1{\sim}5$) were discussed. The ZrO2 jets with 5kg mass and 5cm diameter were poured into the 67cm deep water pool at $30{\sim}95^{\circ}C$. Either spontaneous steam explosions or quenching was observed. The morphology of debris and pressure wave profiles clearly indicates the each case.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.74-78
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• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.1149-1149
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• 2006

The leakage characteristic is an important factor in power plant. However, most of power plant have efficiency problem which is occurred leaking between high pressure steam turbine axle and stator. The labyrinth seal which is used between the main turbine axle and stator in the power plant. Because it is able to be non-contact seal and it is minimize clearance to decrease the leakage. But its actual system is too huge to experiment. Therefore, most steam turbine seal performance tests were conducted by air similarity test. This paper described a test facility and program for air similarity test of high pressure steam turbine seal. A test facility has been designed and built to evaluate leakage verification of labyrinth seal. The test facility consist of air compressor, anti-swirl labyrinth seal for 1/3 air similarity model, pressure transducer, air flow measure system, instrumentation and auxiliary system. For evaluation of steam turbine seal performance, the air similarity test of labyrinth seal leakage verification was conducted and we compared experiment data and analysis result.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.54-60
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• 2017

When a sudden rupture occurs in high energy lines, ejection of inner fluid with high temperature and pressure causes blast wave as well as thrust forces on the ruptured pipe itself. The present study is to examine pipe whip behaviors and blast wave phenomena under postulated pipe break conditions. In this context, typical numerical models were generated by taking a MSL (Main Steam Line) piping, a steam generator and containment building. Subsequently, numerical analyses were carried out by changing break locations; one is pipe whip analyses to assess displacements and stresses of the broken pipe due to the thrust force. The other is blast wave analyses to evaluate the broken pipe due to the blast wave by considering the pipe whip. As a result, the stress value of the steam generator increased by about 7~21% and von Mises stress of steam generator outlet nozzle exceeded the yield strength of the material. In the displacement results, rapid movement of pipe occurred at 0.1 sec due to the blast wave, and the maximum displacement increased by about 2~9%.

• 한국태양에너지학회:학술대회논문집
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• 2008.11a
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• pp.280-287
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• 2008

In case of low pressure steam turbine used in power plant, it was operated in wet steam and high stress condition. Therefore, it is possible that the corrosion damage of low pressure was induced by this condition. According to previous study, about 30% of total blade failure correspond to corrosion fatigue or SCC(stress corrosion cracking) in low pressure turbine. Especially, LSB(last stage bucket) of low pressure turbine has a higher hardness to prevent erosion damage due to water droplet however, generally this is more dangerous for SCC damage. Therefore, to improve reliability of turbine blade. various methods for SCC evaluation has been developed. In this study, the crack found in LSB during in-service inspection was evaluated using microstructure analysis and stress analysis. From the stress analysis, the optimum size of fillet to remove the crack was proposed. And also, the reliability was evaluated for modified LSB using GOODMAN diagram.