• Journal of Energy Engineering
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• v.7 no.2
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• pp.202-208
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• 1998

As in the other industrial processes, a nuclear power plant involves a steam relieving process through which condensable steam is discharged and condensed in a subcooled pool. An analysis of steam discharge transients was carried out using the method of characteristics to determine the flow characteristics and dynamic loads of piping that are used for structural design of the piping and its supports. The analysis included not only the steam flow rate but also the flow rates of the air and water which originally exist in the pipe. The analytical model was developed for a uniform pipe with friction through which the flow was discharged into a suppression pool. Including the combinations of system elements such as reservoir, valve and branching pipe lines. The piping flow characteristics and dynamic loads were calculated by varying system pressure, pipe length, and submergence depth. It was found that the dynamic load, water clearing time and water clearing velocity at the water/air interface were dependent not only on the system pressure and temperature but also on the pipe length and submergence depth.

• Plant Journal
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• v.4 no.3
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• pp.60-65
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• 2008

Power plant industry has been developed at high-capacity, high-technology, and innovation. Steam turbine became the most useful equipment that dominate more than 50% of all the world electricity production. And developed new materials of the turbine blade and extended length of the turbine last blade brought reform in steam turbine performance upgrade. In this paper, when do partial load driving in high-capacity steam turbine, optimum driving method found whether there is something. In operating steam turbine, there is a lot of loss from secondary wake and throttle of the 1st stage nozzle by the biggest leading factor that load fluctuation affects in high-pressure steam turbine performance. Effect of internal efficiency by 1 stage nozzle is the biggest here, but here fluid flow and flow analysis were not yet examined closely definitely. So, Analyzed design data and acceptance performance test result to applying subcritical pressure drum type 560 MW, supercritical-pressure once through type 500 MW, and 800 MW steam turbines actually. In conclusion, at partial load driving, partial arc admission(PAA) is more efficient than full arc admission(FAA) efficiency. This is judged by because increase being proportional with gross energy of stream that is pressure - available energy if pressure of stream that is flowed in to the turbine increases, available energy becomes maximum and turbine efficiency improves. Therefore, turbine performance is that preview that first stage performance fell if decline is serious in partial load because first stage performance changes according to load.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.437-440
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• 2005

In the present work, characteristics of the flow in CAGE of a steam turbine LP/HP Bypass control valve for thermal power plant are investigated. The flow field is analyzed numerically by solving steady three-dimensional Reynolds-averaged Navier-Stokes equations. Shear stress transport (SST) model is used as turbulence closure.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2017.05a
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• pp.637-642
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• 2017

In this study, cold flow and combustion tests are conducted and analyzed to validate designed rocket engine head generating high temperature/pressure steam. At first, uni-injector was designed and manufactured, and cold flow test was conducted. Through this, differential pressure that can supply designed flow rate was confirmed. Also, Each injector's spray pattern were confirmed by patternator. Based on cold flow test results, we selected injectors among the candidates and arranged them on engine head, and cold flow and propellant spray tests were conducted. Finally, combustion test was carried out to analyze the flow rate, pressure, combustion efficiency. As a result, validation of rocket engine head for the development of the high temperature and high pressure steam generator has been completed.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Tribology and Lubricants
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• v.24 no.6
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• pp.362-367
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• 2008

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.889-896
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• 2020

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.13 no.6
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• pp.452-459
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• 2003

The characteristics of fluid-elastic instability and effects of turbulent excitations for the KSNP steam generator tubes were investigated numerically. The information for the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the flow-induced vibration(FIV) analysis has been conducted by using the PIAT(program for Integrity assessment of SG tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the SG. To investigate the cause of the concentrated tube plugging zone, the FIV analysis has been performed for various column and row number of the steam generator tubes. From the results of FIV analysis the stability ratio due to the fluid-elastic instability and vibrational amplitude due to the turbulent excitation in the concentrated plugged zone have a trend of larger values than those of the outer concentrated tube Plugging zone.